| 04-16-2010 | On February 15, 2010, at approximately 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />, a leak was identified on a portion of Emergency Closed Cooling Water subsystem B (ECCW B), American Society of Mechanical Engineers ( ASME) Class 3 piping. On February 16, 2010, at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br />, operators determined that ECCW B was inoperable due to the leak. The leak had been identified approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 40 minutes prior to the inoperability determination; therefore, the plant was in the associated Technical Specification (TS) Limiting Condition for Operation 3.7.10 "ECCW System," Condition B, for longer than the Required Action Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. During this period, ECCW A and its associated supported systems were operable.
The cause of the event was that guidance for evaluating the operability impact of flaws in ASME piping was not adequately captured in plant procedures or training. For the leak itself, fatigue cracking was identified as the primary failure mechanism. The piping leak was repaired. The plant procedure for prompt operability determinations (POD) will be revised to direct consideration of reportability ramifications when establishing POD completion times. Procedure guidance will be developed to address through-wall piping leaks. Training courses will also be revised and Operations as well as other support personnel will be trained to lessons learned from this event.
The safety significance of this event is considered to be low. This event is reported in accordance with 10 CFR 50.73 (a)(2)(i)(B) as an operation or condition which was prohibited by the plant's TS.
0NRC FORM 366 (9-2007) PRINTED ON RECYCLED PAPER |
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LER-2010-002, Piping Leak Results in Condition Prohibited by Technical Specifications| Docket Number |
| Event date: |
02-16-2010 |
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| Report date: |
04-16-2010 |
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| Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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| Initial Reporting |
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| 4402010002R00 - NRC Website |
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Energy Industry Identification System Codes are identified in the text as [XX].
INTRODUCTION
On February 15, 2010, at approximately 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />, a leak was identified on a portion of Emergency Closed Cooling Water [CC] subsystem B (ECCW B), American Society of Mechanical Engineers (ASME) Class 3 piping. On February 16, 2010, at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br />, operators determined that ECCW B was inoperable due to the leak. The leak had been identified approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 40 minutes prior to the inoperability determination; therefore, the plant was in Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.10 "ECCW System," Condition B, for longer than the Required Action B.1 Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. During this period, the plant was operating in MODE 1 (Power Operation) at 100 percent rated thermal power, with ECCW A and its associated supported systems operable. This event is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's TS.
EVENT DESCRIPTION
On February 15, 2010, at approximately 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />, a non-licensed operator performing plant equipment rounds identified a leak in overhead ECCW B piping. The leak was estimated at two to three drops per minute. The leak was located on ASME Class 3, one-half-inch diameter instrument piping for ECCW B, flow instrumentation monitoring ECCW flow through the Control Complex Chilled Water [KM] subsystem B (CCCW B) Chiller. The leak appeared to be at a weld where the piping attached to a 10 inch diameter flow element flange. The leak was reported to the control room.
The Shift Manager, a Senior Reactor Operator (SRO), inspected the leak in the field and reviewed available documentation to evaluate the leak. The Shift Manager completed an Immediate Operability Determination and declared ECCW B "operable but degraded," based on the leakage being minor and that it did not present a challenge to the system's seven day mission time.
On February 16, 2010, at 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br />, the Shift Manager requested a Prompt Operability Determination (POD) from engineering personnel, and assigned a completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Additionally, planning of activities to repair the leak was initiated. Between 0730 and 0815 hours0.00943 days <br />0.226 hours <br />0.00135 weeks <br />3.101075e-4 months <br />, engineering Nondestructive Examination (NDE) personnel inspected the leak site. A crack was seen in the toe of the weld (pipe side) approximately one-third the pipe circumference in length.
Liquid penetrant, magnetic particle, and ultrasonic testing (UT) could not be performed because the piping was leaking water and there was limited access for a UT transducer at the crack site; therefore, the flaw could not be characterized. As a result, engineering personnel met with the Shift Manager and informed him that they could not support a determination of operability for ECCW B.
On February 16, 2010, at 1252 hours0.0145 days <br />0.348 hours <br />0.00207 weeks <br />4.76386e-4 months <br />, the operators entered TS LCO 3.7.4, "Control Room Heating, Ventilation, and Air Conditioning (HVAC)" Condition A (One control room HVAC [VI] subsystem inoperable), when CCCW B was declared inoperable to support isolation and repair activities associated with the ECCW B piping leak.
At 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br />, plant operators declared the ECCW B inoperable, and entered TS LCO 3.7.10, � Condition A (i.e., one ECCW subsystem inoperable). TS LCO 3.7.10, Required Action A.1 was performed to immediately declare associated systems and components inoperable. The equipment declared inoperable included the following:
- Combustible Gas Mixing [BB] subsystem B (TS LCO 3.6.3.3)
The leak was isolated in accordance with plant procedures and at 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br />, TS LCO 3.7.10 CONDITION A was exited when the ECCW B was declared operable. Additionally, TS LCOs 3.5.1, 3.6.1.7, 3.6.2.3, and 3.6.3.3 were exited.
The associated ECCW B instrument piping was replaced and the leak repaired. On February 18, 2010, at 0457 hours0.00529 days <br />0.127 hours <br />7.556217e-4 weeks <br />1.738885e-4 months <br />, TS LCO 3.7.4 was exited when Control Room HVAC subsystem B was declared operable.
CAUSE OF EVENT
The cause of the operation prohibited by TS was that guidance for evaluating the operability impact of flaws in ASME piping was not adequately captured in plant procedures or training. The POD process does not consider reportability ramifications when assigning a completion time.
Consequently, the Shift Manager assigned a completion time without considering the impact if the POD could not support system operability.
For the leak itself, fatigue cracking was identified as the primary failure mechanism. After the crack developed a corrosion component was also involved. During the causal investigation, it was identified that a missing tubing support (i.e., Stauff clamp) downstream of the cracked weld was identified and corrected in 2006. This support piece was most likely missing since plant construction exposing the section of pipe to years of high-cycle, low stress fatigue.
EVENT ANALYSIS
The ECCW system provides a heat sink for the removal of process and operating heat from safety related components during a design basis accident or transient. The ECCW system consists of two independent and redundant closed loop subsystems (A and B). Each subsystem consists of a motor driven pump, heat exchanger, surge tank, and associated piping and valves. The ECCW heat exchangers transfer heat to the Emergency Service Water system. The isolation of ECCW to other components or systems may render those components or systems inoperable, but does not affect the operability of the ECCW system. During the period of ECCW B inoperability, (i.e., from February 15, 2010, at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />, until February 16, 2010, at 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br />), ECCW A and its associated supported systems were operable.
On February 15, 2010, at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />, the plant entered TS LCO 3.7.10, Condition A (One ECCW subsystem inoperable) with a Required Action A.1 to declare associated system(s) or component(s) _NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER � inoperable immediately. The subsystem B inoperability was not recognized, and so the plant entered TS LCO 3.7.10, Condition B (Required Action and Completion Time of Condition A not met) with Required Actions B.1, be in MODE 3 (Hot Shutdown) in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and B.2 , be in MODE 4 (Cold Shutdown) in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. On February 16, 2010, at 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />, TS LCO 3.7.10 Required Action B.1 Completion Time was not met, and a subsequent reportability review determined that this condition was prohibited by the plant's TS and reportable in accordance with 10 CFR 50.73(a)(2)(i)(B). TS LCO 3.7.10 Required Action A.1 was met on February 16, 2010, at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br />, 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 40 minutes after the plant entered Condition A.
The completion times of the additional TS LCOs (i.e., for TS LCOs 3.5.1, 3.6.1.7, 3.6.2.3, 3.6.3.3, and 3.7.4) that were entered due to the ECCW B inoperability, were not challenged. TS LCO 3.5.1 Condition C (i.e., two Emergency Core Cooling System injection subsystems inoperable) has the most limiting Completion Time for Required Action C.1, to restore one injection subsystem to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. TS LCO 3.7.4, for the Control Room HVAC subsystem B (that remained inoperable for the piping repair) Condition A (i.e., one subsystem inoperable) has a Completion Time for Required Action A.1 to restore the subsystem to operable status within 30 days. The actual inoperability time for Control Room HVAC subsystem B due to ECCW B inoperability was approximately two days, seven-and-a-half hours.
The Probabilistic Risk Assessment (PRA) evaluation determined this event to be administrative in nature as a result of noncompliance with TS requirements. The respective function was available.
On this basis, this issue is viewed to be negligible from the PRA perspective.
If a "bounding" evaluation were to be postulated to account for a condition of the piping structural integrity being "indeterminate" from an engineering perspective, and ECCW B was postulated to be unavailable from the time of the leak discovery to the time of isolation, an incremental Conditional Core Damage Probability (ICCDP) of 2.64E-8 was calculated. The results indicate that ICCDP is below 1.0E-6, which is below the guidance threshold for being risk significant. Based on the PRA assessment, this event is considered to be of low safety significance.
CORRECTIVE ACTIONS
The plant procedure for PODs will be revised to direct consideration of reportability ramifications when establishing POD completion times. Additionally, procedure guidance will be developed to address through-wall piping leaks. Training courses will also be revised and Operations and other support personnel trained to the lessons learned from this event. In the interim, written guidance has been provided to the operators.
PREVIOUS SIMILAR EVENTS
The corrective action program database was reviewed for previous similar events occurring in the past three years. LER 2008-002, "Inoperable Emergency Closed Cooling System Results in Condition Prohibited by Technical Specifications" documents that ECCW B was inoperable due to insufficient minimum flow for approximately 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> prior to required actions being taken to declare the associated systems and components inoperable. Identified causes included inadequacies in system operation and clearance procedures. The procedure changes that were developed could not reasonably be expected to prevent the occurrence of this event.
�
COMMITMENTS
There are no regulatory commitments contained in this report. Actions described in this document represent intended or planned actions, are described for the NRC's information, and are not regulatory commitments.
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| | | Reporting criterion |
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| 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 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | | 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 |
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