| 03-25-2010 | On April 27, 2009, the shutdown cooling mode of operation for the Residual Heat Removal ( RHR) system was interrupted for one hour and four minutes. At the time of the event, the plant was in MODE 5 (Refueling). Plant operators verified the designated alternate methods for decay heat removal were available. Reactor coolant temperature increased three degrees F.
The event initiated as Instrumentation and Control technicians were installing an electrical jumper for a surveillance test. During the installation process, the jumper slipped off the terminal connector and created a short circuit to ground which blew the circuit's protective fuse. The blown fuse caused an RHR common suction valve to close (i.e., invalid signal) and trip the operating RHR A subsystem.
Loss of the common suction flow path also prevented the.RHR B subsystem from being operated.
The fuse was replaced and logic reset to open the suction valve and establish RHR B shutdown cooling operation. The individual human performance deficiencies were addressed in accordance with the company's performance management process. Procedure and program changes were made to address weaknesses in risk perception, use of jumpers and lifted leads, and work on protected equipment.
The safety significance of this event is considered to be low. This event is reported in accordance with 10 CFR 50.73 (a)(2)(v)(B) as a condition that could have prevented the fulfillment of the safety function of a system needed to remove residual heat. |
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LER-2010-001, Invalid Isolation Signal Results in Shutdown Cooling Interruption| Docket Number |
| Event date: |
04-27-2009 |
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| Report date: |
03-25-2010 |
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| Reporting criterion: |
10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat |
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| 4402010001R00 - NRC Website |
|
Energy Industry Identification System Codes are identified in the text as [XX].
INTRODUCTION
On April 27, 2009, at 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />, the Residual Heat Removal (RHR) [BO] shutdown cooling outboard common suction isolation valve 1E12-F008 [ISV] received an invalid isolation signal resulting in the operating RHR A pump [P] tripping while the system was in the shutdown cooling mode of operation. When valve 1E12-F008 isolated, both the A and B RHR subsystems became inoperable and shutdown cooling flow to the reactor from the RHR system was interrupted for approximately one hour and four minutes. At the time of the event, the plant was in MODE 5 (Refueling) with the RHR B subsystem in standby. On April 28, 2009, at 0055 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br />, notification was made to the NRC Operations Center (ENF No. 45025) in accordance with 10 CFR 50.72(b)(3)(v)(B), as an event or condition that could have prevented fulfillment of the safety function of a system needed to remove residual heat.
The event notification was retracted on June 25, 2009, based on further evaluation which determined that interrupted operation of the RHR system in the shutdown cooling mode was not reportable in accordance with 10 CFR 50.72 or 50.73 because (1) there was not a reasonable expectation of the loss of the safety function of a system needed for residual heat removal, and; (2) there was no operation or condition prohibited by the plant's Technical Specifications.
The event was later re-evaluated for reportability using additional guidance and enforcement history related to safety system functional failure reporting. The evaluation determined that the April 27, 2009, loss of shutdown cooling event is reportable as a Licensee Event Report (LER) under 10 CFR 50.73(a)(2)(v)(B), "Any event or condition that could have prevented the fulfillment of the safety function of structures, or systems that are needed to remove residual heat." On March 10, 2010 at 1515 hours0.0175 days <br />0.421 hours <br />0.0025 weeks <br />5.764575e-4 months <br />, the NRC Operations
EVENT DESCRIPTION
On April 27, 2009, at 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />, with the plant in MODE 5 during refueling outage 12 and water level less than 22 feet 9 inches above the top of the reactor pressure vessel (RPV) flange, the RHR shutdown cooling outboard common suction isolation valve 1E12-F008 received an invalid isolation signal. This caused the RHR A pump to trip while operating in the shutdown cooling mode of operation. RHR A was the primary (operating) RHR subsystem. RHR B was the backup subsystem as required by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.9.9 "Residual Heat Removal (RHR) — Low Water Level." When valve 1E12-F008 isolated, both the A and B RHR subsystems became inoperable. The capability for shutdown cooling subsystem operation was lost with this configuration. The operators entered Off-Normal Instruction (ONI) E12 2, "Loss of Decay Heat Removal" and TS LCO 3.9.9, Conditions A and C due to closure of valve 1E12-F008. The operators took the appropriate required actions to comply with the LCO requirements including Required Action A.1, to verify that an alternate method of decay heat removal was available for each inoperable RHR shutdown cooling subsystem; Required Action C.1, to verify reactor coolant circulation by an alternate method; and Required Action C.2, to monitor reactor coolant temperature.
0 At 1735 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.601675e-4 months <br />, a blown fuse was discovered in control room panel 1H13-P691. The fuse services the circuit that provides 24 vdc electrical power to Trip Unit Card File & Calibration Unit Z1A. When the fuse blew, the 24 vdc electrical power was lost to Trip Unit 1B21-N679A "Reactor Pressure High." Relay 1B21-K124A de-energized, which caused automatic closure of valve 1E12-F008 and the subsequent automatic tripping of the RHR A pump, as designed.
Plant operators pursued parallel paths (both local and remote) to manually realign the valve for RHR B subsystem operation. At 1816 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.90988e-4 months <br />, the fuse was replaced. The isolation logic initiated by the blown fuse was reset. Valve 1E12-F008 was aligned manually to the open position from the control room to support RHR B subsystem operation. At 1834 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.97837e-4 months <br />, the RHR B pump was started to restore shutdown cooling operation in compliance with TS LCO 3.9.9. The amount of time that RHR shutdown cooling was interrupted was one hour and four minutes. During that time, reactor coolant temperature increased three degrees F from 94 to 97 degrees F.
On April 28, 2009, after performing fill and vent activities on the RHR A subsystem, at 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />, the RHR A subsystem was declared operable and TS LCO 3.9.9, Condition A was exited. At 0458 hours0.0053 days <br />0.127 hours <br />7.572751e-4 weeks <br />1.74269e-4 months <br />, the RHR A subsystem was started in the normal shutdown cooling lineup and at 0513 hours0.00594 days <br />0.143 hours <br />8.482143e-4 weeks <br />1.951965e-4 months <br />, the RHR B subsystem was placed in standby.
CAUSE OF EVENT
The interruption of RHR shutdown cooling system operation was caused by a blown fuse which sent an isolation signal to valve 1E12-F008. Closure of this valve sent an automatic trip signal to the operating RHR A subsystem and, with the valve now closed, prevented immediate startup of the RHR B pump to restore shutdown cooling.
The fuse blew while Instrumentation and Control (I&C) technicians were installing a jumper in control room panel 1H13-P691 as part of prerequisites for performing the containment integrated leak test (ILRT). While the l&C technicians attempted to attach the jumper wire, the mini-grabber on the jumper slipped off the terminal connector, made contact with the ground bus, created a short circuit to ground and blew the fuse. This represented a human performance error by the l&C technicians.
A root cause evaluation performed for this event identified the following causes which established the conditions leading to the improper jumper installation:
- Organizational and individual weaknesses exist with risk perception and mitigation. The pre-job briefing for the ILRT prerequisite task did not include a specific discussion of risk and risk mitigation. The inherent risk of using jumpers, specifically in the installed location, was not recognized and mitigated. Planning/review/scheduling of outage work activities did not include adequate risk determination/risk management.
- There was continued tolerance and use of less than adequate tools (i.e., procedures, labeling, jumpers, etc) needed for successful task performance. Difficult installation of jumpers and lifted leads had become a routine and accepted practice due to the frequency of use in past work activities and procedures.
�
- Work permitted on or near protected equipment is outside industry norms. Control Room panel 1H13-P691 was posted as protected equipment.
- Use of jumpers, lifted leads, and difficult measuring and test equipment (M&TE) connections had not been rigorously challenged and corrected.
EVENT ANALYSIS
The purpose of the RHR system in MODE 5 is to remove decay heat and sensible heat from the reactor coolant. There are two redundant, manually controlled shutdown cooling subsystems (RHR A/B) available to perform the shutdown cooling function. Each subsystem loop consists of one motor driven pump, two heat exchangers in series, and associated piping and valves. Both subsystems share a common suction from the same recirculation loop. Motor operated valves 1E12-F008 and 1E12-F009 are located in this line to provide inboard and outboard containment isolation. Each pump discharges coolant to the reactor after it has been cooled by circulation through the respective heat exchangers. The RHR heat exchangers transfer heat to the Emergency Service Water System.
In MODE 5, decay heat removal by the RHR system in the shutdown cooling mode is not required for mitigation of any events or accidents evaluated in the safety analyses.
For this event, shutdown cooling was lost for one hour and four minutes in MODE 5. During that time, reactor coolant temperature rose from 94 to 97 degrees F. The reactor coolant time-to-boil was approximately nine hours due to low decay heat level in the core.
The operators promptly verified the alternate methods of decay heat removal were available for the inoperable RHR A/B shutdown cooling subsystems in accordance with TS 3.9.9, Required Action A.1 A qualitative probabilistic risk assessment (PRA) was performed for the duration that shutdown cooling was interrupted. Based on the availability of the Shutdown Cooling function, the timeframe involved before boiling of the reactor vessel inventory was expected, and the other mitigating alignments available to preclude reaching the boiling point had they been required, this event is considered as having a low safety significance. The conclusion of Shutdown Cooling availability is in alignment with the guidance/definition provided in NRC Inspection Manual 0609, Significance Determination Process, Appendix G, Shutdown Operations.
CORRECTIVE ACTIONS
The remaining refueling outage activities were reviewed jointly by Operations and Maintenance to drive improved execution of risk-significant work. The risks were defined and appropriate measures were developed to manage those risks. Among the actions taken were:
- Sessions were conducted with Maintenance and Operations personnel, reinforcing expectations to consistently use Human Performance tools and behaviors;
- Identifying and reviewing maintenance activities that could affect shutdown cooling, reactor water level or pressure control, or reactivity management control and developing mitigation strategies for maintenance items identified as risk-significant; and _ NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER
- Flagging work packages with yellow, orange, or red elevated work authorization forms.
A Human. Performance Strategic Plan was developed to establish actions to improve:
- Organizational and individual risk perception and mitigation;
- The rigor and intrusiveness for risk assessment of daily work activities; and
- Site culture with respect to continued tolerance and use of less than adequate tools (procedures/work packages, jumpers, and labels) needed for successful task performance.
The protected equipment process established in Nuclear Operating Instruction (NOP)-0P-1007, "Risk Management," was revised to strengthen the controls, access requirements and limitations for work on protected equipment.
A project plan was created to establish the necessary strategies to identify, prioritize, and implement engineered solutions in order to eliminate the use of difficult jumpers, lifted leads, and M&TE connections. Potential solutions include the use of alternate locations or use of other solutions such as test lugs, test switches, sliding links, test boxes, or use of a robust barrier.
PREVIOUS SIMILAR EVENTS
A review of Perry LERs and the corrective action program database for the past three years found one instance of a loss of shutdown cooling event. On July 11, 2007, with the plant in MODE 4 (Cold Shutdown), the RHR B pump tripped off while operating in shutdown cooling. The pump trip occurred when an l&C technician performing a Reactor Core Isolation. Cooling (RCIC) test unnecessarily loosened a wire connection from an electrical terminal, inducing a current from the RCIC circuitry into the electrically independent RHR B trip system. Plant modifications to separate the wiring and install noise suppression diodes were initiated. This event was reported under LER 2007-002.
Corrective actions for the July 11, 2007, loss of shutdown cooling event were directed toward fixing a latent vendor design deficiency and would not have prevented the April 27, 2009, loss of shutdown cooling event. A common factor in both events, however, is that they were initiated by an l&C human performance error. The individual human performance shortfalls were addressed in accordance with the company's performance management process.
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 |
|---|
| 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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