Unit Status: At the time of the event, Units 1 and 2 were in Mode 1 at 100% power.
Event Description:F From approximately 1907 hours0.0221 days <br />0.53 hours <br />0.00315 weeks <br />7.256135e-4 months <br /> to 1938 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.37409e-4 months <br /> on January 19, 2010, the Control Room received intermittent annunciators and bi-stable indications.
Investigation determined this was due to power cycling off and on to some loads served by 120VAC Vital I&C panel board 2EKVA.
W One load affected was Channel 1 of Train A of the Unit 2 Solid State Protection System (SSPS).WWhile power was removed from this Channel, Train A would have been incapable of generating an Engineered Safety Features Actuation System signal. Concurrent with the cycling of power to Channel 1 of Train A of Unit 2 SSPS, the 2B Train of Containment Spray (NS)W were inoperable for scheduled maintenance. and the 2B Diesel Generator (DG)W The above conditions could have delayed initiation of the Unit 2 safety injection DG actuation and the Unit 2 NS safety functions. McGuire is reporting this as a condition that could have prevented. fulfillment of a safety function needed to mitigate the consequences of an accident.
Event Cause: Fabrication deficiency caused a loose bus bar connection in panel board, 2EKVA which rendered Train A of Unit 2 SSPS inoperable concurrent with maintenance on the redundant train.
Corrective Actions: Moved Unit 2 SSPS Train A Channel 1 power to another 2EKVA breaker. Maintenance procedure for Unit 1 & 2 120VAC/125VDC Vital I&C panel boards was revised to ensure these panels are adequately inspected for loose bus bar connections.W Inspections of panel connections are planned and, based upon results, replacement of these panel boards will be considered.
NRC FORM 36e(7-2001) |
LER-2010-001, Loose connection in a panel board serving a Solid State Protection System Train concurrent with redundant train maintenance could have prevented fulfillment of a safety function.| Docket Number |
| Event date: |
|
|---|
| Report date: |
|
|---|
| Reporting criterion: |
10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
|---|
| Initial Reporting |
|---|
| 3702010001R00 - NRC Website |
|
BACKGROUND
The following information is provided to assist readers in understanding the event described in this LER. Applicable Energy Industry Identification [EIIS] system and component codes are enclosed within brackets. McGuire unique system and component identifiers are contained within parentheses.
The Solid State Protection System [JC](SSPS) consists of two trains (A and B), each consisting of four input channels (I, II, III, and IV), a logic train, and an output train. In response to SSPS input signals satisfying the applicable SSPS logic, slave relays in SSPS will actuate as needed a reactor trip and the following safety functions via an Engineered Safety Features Actuation System [JE](ESFAS) signal:
- Phase A and Phase B Containment Isolation
- . Containment Ventilation Isolation.
- . . Steam Lifte Isolation.
- Turbine Generator Trip.. .
- Automatic Switchover of Emergency Core Cooling Water to the Containment Sump
- Containment Air Return and Hydrogen Skimmer Fans SSPS also provides Control Room annunciator, status light, and computer input signals which indicate the condition of bistable input signals, partial trip and full trip functions and the status of the various blocking, permissive and actuation functions.
Each SSPS Train is powered through fuses from the appropriate 120 VAC Vital I&C [EF](EPG) panel board. The EPG panel board supply for Channel 1 of SSPS Train A provides power for the A Train slave relays. The EPG panel board supply for Channel 4 of SSPS Train B provides the power for the B Train slave relays.
EVENT DESCRIPTION
On January 15, 2010 at approximately 1629 hours0.0189 days <br />0.453 hours <br />0.00269 weeks <br />6.198345e-4 months <br />, the Control Room received ten annunciators and bi-stable lights. These indications cycled on and off for a period. of approximately ten minutes and then remained off. During the period of time these indications were cycling on and off, no abnormal plant or equipment transients occurred. A failure investigation team was established to perform troubleshooting, evaluate plausible failure modes, and determine the failure cause.
On January 19, 2010 at approximately 1907 hours0.0221 days <br />0.53 hours <br />0.00315 weeks <br />7.256135e-4 months <br />, during troubleshooting related to the January 15, 2010 indications, the Control Room received twelve annunciators and bi-stable lights. These indications cycled on and off until approximately 1938 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.37409e-4 months <br /> and then remained off. Ten of these annunciators and bi-stable lights were identical to the indications received on January 15, 2010. During the period of time these indications were cycling on and off, no abnormal plant or equipment transients occurred.
On January 20, 2010, based upon evidence gathered to date, including the fact that the annunciators and bi-stable lights were associated with circuits receiving power from Unit 2 EPG panel. board 2EKVA, the failure investigation team measured voltages and currents, performed thermography scans, and implemented visual inspections of panel board 2EKVA. Although these activities did not identify any abnormalities, they did identify a common bus bar connection on both the line side and the neutral side of panel board 2EKVA breaker8 3 and 4. 'A thorough'inspectioh of this connection was hampered since these breakers were energized and the connection was difficult to see.
On January 21, 2010, a critical activity plan was approved which provided for isolating power to Unit 2 EPG panel board 2EKVA breakers 3 and 4 to facilitate a thorough inspection of the common bus bar connection identified on January 20, 2010. As part of this inspection, this connection was to be checked for tightness.
On January 22, 2010 at 1312 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.99216e-4 months <br />, while preparing to implement the critical activity plan, the Control Room received multiple annunciators and bi-stable lights similar to those received on January 15, 2010 and January 19, 2010.
These indications cycled on and off until approximately 1324 hours0.0153 days <br />0.368 hours <br />0.00219 weeks <br />5.03782e-4 months <br />. During the period of time these indications were cycling on and off, no abnormal plant or equipment transients occurred.
On January 22, 2010, while performing inspections of Unit 2 EPG panel board 2EKVA breakers 3 and 4 as per the critical activity plan, it was observed that the neutral connection under breaker 4, between the main bus bar and the bus connector piece for breakers 3 and 4 was significantly loose. The screw associated with this connection had backed out to the point that the bus bar connector piece was no longer making firm contact with the main bus bar.
On January 22, 2010, a modification was implemented which moved the electrical power cables previously connected to Unit 2 EPG panel board 2EKVA breakers 3 and 4 to previously spare breakers 24 and 9 on panel board 2EKVA.
On January 25, 2010, a reportability evaluation was initiated after investigation determined the intermittent annunciators and bi-stable indications received on January 15, 2010, January 19, 2010, and January 22, 2010 occurred concurrent with power to Channel 1 of Train A of Unit 2 SSPS cycling off and on as a result of the loose bus bar connection common to Unit 2 EPG panel board breakers 3 and 4.
On January 26, 2010, it was discovered that while powerto_Channel 1 of Train A of Unit 2 SSPS was cycled off,.none of the slave relays associated With Train A would .have been. capable of generating an ESFAS signal if needed; which would haVe prevented initiation of the Unit 2 Train A ESFAS related Safety functions.
On January 26, 2010, a review was performed to identify Unit 2 Train B equipment that was inoperable during the periods of time on January 15, 2010, January 19, 2010, and January 22, 2010 when power to Channel 1 of Train A of Unit 2 SSPS was cycling off and on. This review identified that, due to scheduled maintenance, the Unit 2 NS Train B safety function and the Unit 2 Train B safety injection DG actuation safety function were inoperable during the period of time on January 19, 2010 when power to Channel 1 of Train A of Unit 2 SSPS was cycling off and on (approximately 1907 hours0.0221 days <br />0.53 hours <br />0.00315 weeks <br />7.256135e-4 months <br /> to 1938 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.37409e-4 months <br />). During this period of time, if SSPS had attempted to initiate the Unit 2 NS safety function and the Unit 2 safety injection DG actuation safety function while Channel 1 of Train A power was cycled off, fulfillment of these safety functions would have been delayed for a period of time that would not support the assumptions and conclusions of the applicable safety analyses. This represented a condition that could have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident, which is reportable as per the requirements of 10 CFR 50.73(a)(2)(v)(D). Note, when power to Channel 1 cycled back on, SSPS contacts associated with the Unit 2 NS Train A safety function and the Unit 2 Train A safety injection DG actuation safety function would have latched in and these functions would have initiated. Upon latch in, these Train A safety functions would have continued as needed even if power to Channel 1 of Train A of Unit 2 SSPS had cycled back off.
CAUSAL FACTORS
The cause of the intermittent loss of power to Channel 1 of Train A of Unit 2 SSPS was a manufacturer fabrication deficiency. Specifically, a degraded connection under EPG panel board 2EKVA breaker 4, between the main bus bar and the breaker to bus bar connector for breakers 3 and 4, had developed an intermittent loose connection. This loose connection occurred when the connecting screw that holds the bus connector piece to the main bus bar backed out of the threaded hole in the main bus bar. Either the screw was not tightened properly initially or it was not designed/fabricated with sufficient locking mechanisms to prevent it from backing out over time.
This connection is inaccessible without removing the breakers and it is difficult to see. This screw loosened to the point where the electrical continuity of the bus bar connection was lost, thereby causing a loss of power to Channel 1 of Train A of Unit 2 SSPS. Upon losing electrical continuity, the increased resistance across the degraded bus bar connection caused temperature to increase, restoring the electrical continuity of the SSPS. Subsequent load fluctuations resulted in a decrease in resistance across the bus bar connection causing the temperature of the connection to drop causing the bus bar connection to lose electrical continuity, again causing a loss of power to Channel 1 of Train A of Unit 2 SSPS. This cyclic loss and restoration of electrical continuity at the bus bar connection for EPG panel board 2EKVA breaker 4 resulted in the intermittent loss of power to Channel 1 of Train A of Unit 2 SSPS.
Contributing Causes:
A contributing cause to this event was inadequate preventive maintenance on Unit 2 EPG panel board 2EKVA. A preventive maintenance (PM) procedure for the panel board provided for periodic inspection of the panel board for loose connections. However, the PM procedure did not provide specific inspection instructions for connections that were difficult to see, or not shown on the applicable vendor documents, as in the case of the degraded bus bar connection common to breakers 3 and 4 of EPG panel board 2EKVA.
Therefore, whenever the PM was performed on panel board 2EKVA, the connection common to breakers 3 and 4, which was difficult to see, was not inspected for a loose connection.
CORRECTIVE ACTIONS
Immediate:
1. Implemented a modification which moved the power supply for Channel 1 of Train A of Unit 2 SSPS from breaker 4 on Unit 2 EPG panel board 2EKVA to previously spare breaker 9 on that panel board. As part of this modification, the tightness of the breaker 9 bus bar connections was verified prior to placing the breaker back in service.
Subsequent:
1. The PM procedure which provides, for periodic inspection of the Unit 1 & 2 120VAC/125VDC Vital I&C panel boards was revised to ensure bus bar connections (including those bus bar connections that are inaccessible, difficult to see, or not shown on the applicable vendor documents) are adequately checked for loose connections.
Planned:
1.Starting with Unit 1 outage 1E0C20 (currently in progress)and the next Unit 2 outage (2E0C20), inspections of bus bar connections on the Unit 1 & 2 120VAC/125VDC Vital I&C panel boards for loose connections are planned. To date, 1E0C20 inspections of panel board bus bar connections have not identified any findings of significance.
2.Based upon results of the planned bus bar inspections of the Unit 1 & 2 120VAC/125VDC Vital I&C panel boards, modifications to replace these panel boards will be considered. If these modifications are implemented, bus bar connections in the new panel boards will be designed and fabricated to prevent loosening, or designed for better accessibility or viewing.
SAFETY ANALYSIS
safety injection DG actuation safety function while Channel 1 of Train A of Unit 2 SSPS power was cycled off, fulfillment of these safety functions would have been delayed for a period of time which would not support the assumptions and conclusions of the applicable safety analyses. Note, when NS Train A safety function and the Unit 2 Train A safety injection DG actuation safety function would have latched in and these functions would have initiated. Upon latch in, these Train A safety functions would have continued as needed even if power to Channel 1 of Train A of Unit 2 SSPS cycled back off. The longest period of time that power to Channel 1 of Train A of Unit 2 SSPS was cycled off between 1907 hours0.0221 days <br />0.53 hours <br />0.00315 weeks <br />7.256135e-4 months <br /> and 1938 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.37409e-4 months <br /> on January 19, 2010 was approximately 2.25 minutes. This represents the longest duration of time on that date where fulfillment of the Unit 2 NS safety function and the'Unit 2 safety injection DG actuation safety function would not have occurred if needed.
A risk-informed approach was used to determine the significance associated with unavailability of Train A of Unit 2 SSPS and coincident unavailability of the Unit 2 NS Train B and the Unit 2 B Train DG. Conservatively, the unavailability duration used in this analysis was assumed to be 31 minutes (from approximately 1907 hours0.0221 days <br />0.53 hours <br />0.00315 weeks <br />7.256135e-4 months <br /> to 1938 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.37409e-4 months <br /> on January 19, 2010). Due to the short duration of unavailability of Train A of Unit 2 SSPS, the Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP) associated with this event are negligible, based on being less than 1E-8 and 1E-9, respectively.
Given the above, this event is considered to be of no significance with respect to the health and safety of the public.
ADDITIONAL INFORMATION
To determine if a recurring or similar event exists, a search of the McGuire Problem Identification Process (PIP) database was conducted for a time period covering 5 years prior to the date of this event. Based on Duke's definition of a recurring event, similar significant event with the same cause code, no recurring events were identified.
|
|---|
|
|
| | | Reporting criterion |
|---|
| 05000220/LER-2010-001 | Reactor Scram Due to Inadequate Post Maintenance Testing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000410/LER-2010-001 | Reactor Scram Due to Inadvertent Actuation of the Redundant Reactivity Control System During Maintenance | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000395/LER-2010-001 | Reactor Building Cooling Units Reduced Air Flow Rate Below Technical Specification Limits | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000382/LER-2010-001 | Spent Fuel Pool Cooling Single Failure | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(ii) | | 05000374/LER-2010-001 | High Pressure Core Spray System Declared Inoperable Due to Failed Room Ventilation Control Relay | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | | 05000373/LER-2010-001 | Unauthorized Individual Gained Access to the Protected Area. | | | 05000370/LER-2010-001 | Loose connection in a panel board serving a Solid State Protection System Train concurrent with redundant train maintenance could have prevented fulfillment of a safety function. | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | | 05000261/LER-2010-001 | Emergency Diesel Generator Inoperable in Excess of Technical Specifications Allowed Completion Time | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000263/LER-2010-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000255/LER-2010-001 | Potential Loss of Safety Function Due to a Service Water Pump Shaft Coupling Failure | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000266/LER-2010-001 | Engineered Safety Features Steam Line Pressure Dynamics Modules Discovered Outside of Technical Specification Values | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000306/LER-2010-001 | Unit 2 Turbine Trip during Reactor Shutdown Resulting in a Reactor Scram | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000305/LER-2010-001 | Safety Injection Pump Recirculation Line Isolation Results in Violation of Technical Specifications | | | 05000298/LER-2010-001 | Cooper Nuclear Station 05000298 1 of 4 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000269/LER-2010-001 | Standby Shutdown Facility Letdown Line Orifice Strainer Blocked by Valve Gasket Material | 10 CFR 50.73(a)(2)(i)(b) | | 05000282/LER-2010-001 | Unanalyzed Condition Due to Postulated High Energy Line Break On Cooling Water System | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000277/LER-2010-001 | Multiple Slow Control Rods Results in Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i) | | 05000361/LER-2010-001 | Broken Manual Valve Prevents Timely Condensate Storage Tank Isolation | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000483/LER-2010-001 | Emergency Core Cooling System MODE 4 Operating Practices Prohibited by current Technical Specification 3.5.3 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000498/LER-2010-001 | Unit Shutdown Required by Technical Specifications | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | | 05000316/LER-2010-001 | Valid Actuation of Auxiliary Feedwater System in Response to Valid Steam Generator Low-Low Levels | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000321/LER-2010-001 | Corrosion Induced Bonding Results in Safety Relief Valve Lift Setpoint Drift | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000336/LER-2010-001 | Millstone Power Station Unit 2 Reactor Trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000413/LER-2010-001 | Technical Specification Violation Associated with Failure to Perform Offsite Circuit Verification | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000440/LER-2010-001 | Invalid Isolation Signal Results in Shutdown Cooling Interruption | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | | 05000424/LER-2010-001 | Breaker Failure Results in I B Train Containment Cooling System Being Declared Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000416/LER-2010-001 | Automatic Reactor Scram On Decreasing Reactor Water Level Due To Inadvertent Reactor Feed Pump Trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000249/LER-2010-001 | OPRM Power Supply Failure during Maintenance Results in Unit 3 Automatic Reactor Scram | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000251/LER-2010-001 | Two Shutdown Bank Rods Were Dropped from Fully Withdrawn Position | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(iv), System Actuation | | 05000261/LER-2010-002 | Plant Trip due to Electrical Fault | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000255/LER-2010-002 | Condition that Could Have Prevented the Fulfillment of a Safety Function | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000335/LER-2010-002 | Opened ECCS Boundary Door in Violation of Identified Compensatory Measures | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000334/LER-2010-002 | 270 Degree Circumferential Flaw Found on Residual Heat Removal System Drain Valve Socket Weld | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000316/LER-2010-002 | Containment Divider Barrier Seal Mounting Bolts Not Properly Installed | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000250/LER-2010-002 | Fuel Transfer Pump Failure Renders 3B Emergency Diesel Generator Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000269/LER-2010-002 | Manual Reactor Trip due to 1A1 and 1A2 Reactor Coolant PumDHigh Vibration Indication | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000315/LER-2010-002 | Manual Auxiliary Feedwater Actuation in Response to Main Feedpump Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000271/LER-2010-002 | Inoperability of Main Steam Safety Relief Valves due to Degraded Thread Seals | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000277/LER-2010-002 | Improperly Fastened Rod Hanger Results in Inoperable Subsystem of Emergency Service Water | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000413/LER-2010-002 | Discovery of Reactor Coolant System Pressure Boundary Leak at Thermowell 1NCTW5850 Seal Weld. | | | 05000282/LER-2010-002 | Postulated Flooding of Unit 1 Fuel Oil Transfer Pump Motor Starters Could Have Resulted In Reduced Fuel Oil Inventory | | | 05000414/LER-2010-002 | Duke Energy Corporation
Catawba Nuclear Station
4800 Concord Road
York, SC 29745
803-701-4251
803-701-3221 fax
December 15, 2010
U.S. Nuclear Regulatory Commission
Attention: Document Control Desk
Washington, D.C. 20555
Subject:�Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Unit 2
Docket No. 50-414
Licensee Event Report 414/2010-002
Pursuant to 10 CFR 50.73(a)(1) and (d), attached is Licensee Event Report 414/2010-002,
Revision 0 entitled, "Technical Specification Violation Involving Mode Change with Inoperable
Auxiliary Feedwater System Train Due to Closed Pump Discharge Valves".
This report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B).
There are no regulatory commitments contained in this letter or its attachment.
This event is considered to be of no significance with respect to the health and safety of the
public.
If there are any questions on this report, please contact L.J. Rudy at (803) 701-3084.
Sincerely,
faius4- A
James R. Morris
LJR/s
Attachment
www.duke-energy.corn (14
Document Control Desk
Page 2
December 15, 2010
xc (with attachment):
L.A. Reyes
Regional Administrator
U.S. Nuclear Regulatory Commission - Region II
Marquis One Tower
245 Peachtree Center Ave., NE Suite 1200
Atlanta, GA 30303-1257
J.H. Thompson (addressee only)
NRC Project Manager
U.S. Nuclear Regulatory Commission
Mail Stop 8-G9A
11555 Rockville Pike
Rockville, MD 20852-2738
G.A. Hutto, Ill
NRC Senior Resident Inspector
Catawba Nuclear Station
INPO Records Center
700 Galleria Place
Atlanta, GA 30339-5957
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013
(10-2010) Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported
lessons learned are incorporated into the licensing process and fed back to industry. Send commentsLICENSEE EVENT REPORT (LER) regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F53), U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mail to
infocollectssesource@nre.gov, and to the Desk Officer, Office of Information and Regulatory Affairs,
NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used(See reverse for required number of to impose an information collection does not display a currently valid OMB control number, the NRCdigits/characters for each block)
may not conduct or sponsor, and a person is not required to respond to, the info(mation
collection.
1.. FACILITY NAME 2. DOCKET NUMBER I3. PAGE
Catawba Nuclear Station, Unit 2 05000414 1 OF 7
4. TITLE
Technical Specification Violation Involving Mode Change with Inoperable Auxiliary Feedwater System Train Due
to Closed Pump Discharge ValvesD • | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000306/LER-2010-002 | Unit 2 Turbine Shutdown Due To the Loss of a Main Feed Water Pump That Resulted in a Reactor Scram | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000440/LER-2010-002 | Piping Leak Results in Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000382/LER-2010-002 | Main Feedwater Isolation Valve B exceeded allowed outage time due to tubing connection failure | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000370/LER-2010-002 | ref Energy®
REGIS T. REPKO
Vice President
McGuire Nuclear Station
Duke Energy
MGO1VP / 12700 Hagers Ferry Rd.
Huntersville, NC 28078
980-875-4111
980-875-4809 fax
regis.repko(Codu ke-energy.corn
10 CFR 50.73
May 10, 2011
U.S. Nuclear Regulatory Commission
ATTENTION: Document Control Desk
Washington, D.C. 20555
Subject:
D
Duke Energy Carolinas, LLC
McGuire Nuclear Station, Unit 2
Docket Nos. 50-370
Licensee Event Report (LER) 370/2010-02, Supplement 1
Problem Investigation Process (PIP) M-10-05982
Pursuant to 10 CFR 50.73 Sections (a) (1) and (d), attached is Supplement 1 to
Licensee Event Report 370/2010-02, regarding past inoperability of the Unit 2 "A" Train
Nuclear Service Water System and satisfies the commitment to supplement the LER
following completion of the root cause analysis
This supplement to LER 370/2010-02 supersedes the LER previously submitted
December 20, 2010. Completion of the root cause analysis has not affected the original
reporting criteria which was completed in accordance with 10 CFR 50.73 (a) (2) (i) (B),
an Operation Prohibited by Technical Specifications, and 10 CFR 50.73 (a) (2) (v) (B),
any Event or Condition That Could Have Prevented Fulfillment of the Safety Function
needed to remove residual heat.
Additionally, the supplement did not affect the significance of the event which was
considered to be of no significance with respect to the health and safety of the public.
There are no regulatory commitments contained in this report.
If questions arise regarding this LER, contact Rick Abbott at 980-875-4685.
Very truly yours,
Zi1:77
Regis T. Repko
Attachment
www. duke-energy. corn
U.S. Nuclear Regulatory Commission
May 10, 2011
Page 2
cc:�V. M. McCree, Regional Administrator
U.S. Nuclear Regulatory Commission, Region II
Marquis One Tower
245 Peachtree Center Ave., NC, Suite 1200
Atlanta, Georgia 30303-1257
Jon H. Thompson (Addressee Only)
Senior Project Manager (McGuire)
U.S. Nuclear Regulatory Commission
11555 Rockville Pike
Rockville, MD 20852-2738
J. B. Brady
Senior Resident Inspector
U.S. Nuclear Regulatory Commission
McGuire Nuclear Station
W. L. Cox Ill, Section Chief
North Carolina Department of Environment and Natural Resources
Division of Environmental Health
Radiation Protection Section
1645 Mail Service Center
Raleigh, NC 27699-1645
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB. NO 3150-0104 EXPIRES: 08/31/2013
(10-2010) Estimated burden per response to comply with this mandatory collection request: SO hours.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA/Privacy Section (T-5 F53), U.S.
Nuclear Regulatory Commission. Washington, DC 20555-0001, or by Internet e-mail to info
(See reverse for required number of collects resmirceOnrc.gov, and to the Desk Officer, Office of Information and Regulatory
digits/characters for each block) Affairs, NEOB-10202, (3150-01041, Office of Management and Budget, Washington, DC
20503. If a means used to impose an information collection does not display a currently valid
OMB control number, the NRC may not conduct or sponsor, and a person is not required
to respond to, the information collection.
LICENSEE EVENT REPORT (LER)
1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE
McGuire Nuclear Station,2Unit 2 05000-212
0370 OF-7
4. TITLE
Unit 2 Nuclear Service Water System "A" Train Past Inoperable due to
Failed Strainer Differential Pressure Instrument. | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000369/LER-2010-002 | | | | 05000456/LER-2010-002 | Limiting Condition for Operation Action Not Completed Within the Required Time | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000249/LER-2010-003 | Steam Leak Results in HPCI Inoperability | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | | 05000251/LER-2010-003 | Damaged Speed Sensor Caused the 4A Emergency Diesel Generator to be Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
|