03-14-2008 | On January 14, 2008, the plant was operating at approximately 100 percent power with the Reactor Core Isolation Cooling system ( RCIC) considered operable and in standby readiness.
Plant staff performed a review of the RCIC flow controller output on a computer point in preparation for planned maintenance. The flow controller output, monitored on a computer point, had changed from approximately 102.9 percent to approximately 100.2 percent without an observed change on the output meter.
Further review of the flow controller output history identified additional instances of similar changes. Based on this information the RCIC system was determined to have been inoperable for about 35 days. Technical Specification Limiting Condition for Operation 3.5.3 was not met since the condition had been previously unrecognized. This condition is reportable as a Condition prohibited by Technical Specifications, 10CFR50.73(a)(2)(i)(B).
The most probable cause of the observed flow controller anomaly was identified as an intermittent failure of the connector. The Bailey 701 flow controller and connecter, the power supply, the ramp generator/signal converter, and the computer input circuit board were replaced. Planning is in process to replace the obsolete Bailey flow controllers. Additional changes will be made to improve system reliability through improved monitoring of the flow controller output. |
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LER-2008-001, ' Condition Prohibited by Technical Specifications Due to Unrecognized Reactor Core Isolation Cooling InoperabilityDocket Number |
Event date: |
01-14-2008 |
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Report date: |
03-14-2008 |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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4402008001R00 - NRC Website |
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PRINTED ON RECYCLED PAPER
NRC FORM 366A� (9-2007) U.S. NUCLEAR REGULATORY COMMISSION
CONTINUATION SHEET
Energy Industry Identification System Codes are identified in the text as [XX]
INTRODUCTION
On January 14, 2008, at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, the Reactor Core Isolation Cooling system (RCIC) [BN] was declared inoperable due to flow controller [TC] instability as observed on a computer point. At the time of the event, the plant was operating at approximately 100 percent of rated thermal power. The RCIC system was in standby readiness and all Emergency Core Cooling Systems were operable. Subsequent investigations determined that similar instability had occurred on the computer point dating back to December 10, 2007 (i.e. 35 days).
EVENT DESCRIPTION
On January 14, 2008, the plant was operating at approximately 100 percent power with the Reactor Core Isolation Cooling system (RCIC) in standby readiness. Plant operating staff performed a review of the RCIC Bailey 701 flow controller output signal on a computer point in preparation for a planned maintenance outage for the motor-driven feed water pump. The indicated flow controller output had changed from approximately 102.9 percent to approximately 100.2 percent, the normal value, without an observed change on the flow controller output meter. The output signal had shifted from 100.2 percent to 102.9 percent just prior to the condition being identified. Flow controller output should remain steady at 100 percent when in standby readiness. Following review of this information, the RCIC system was declared inoperable at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> due to the flow controller instability on January 14, 2008. This condition is reportable as a Condition prohibited by Technical Specifications, 10CFR50.73(a)(2)(i)(B) A problem solving team was assembled to address this condition. During the review of the flow controller output history, additional instances of similar changes dating back to December 10, 2007 were identified. Based on this information the RCIC system was determined to have been inoperable for about 35 days.
Efforts to isolate the cause of the flow controller anomaly were complicated since the flow controller output had returned to the expected value at the time of discovery and remained there during the troubleshooting activities. The problem solving team developed a failure modes and effects problem solving plan and methodically investigated each component that could result in the output fluctuations. When all in-place testing was completed without identifying the hardware cause, plant management made the decision to remove all potentially contributing components and replace them with new or refurbished components.
The replaced components included the flow controller and connecter, the power supply [RJX], the ramp generator/signal converter [CNV], and the computer input circuit board.
The removed components were sent to the FirstEnergy Beta Lab for further testing. All were tested by the lab with no distinct failure noted.
Following replacement of the above components, the RCIC system was monitored, retested successfully and returned to service on January 21, 2008, at 0454 hours0.00525 days <br />0.126 hours <br />7.506614e-4 weeks <br />1.72747e-4 months <br />.
� Additional monitoring of the flow controller output since that time has not identified any additional occurrences of the flow controller output anomaly.
CAUSE OF EVENT
The flow controller output change is likely the result of an intermittent failure of one of the four replaced components (flow controller and connecter, the power supply, the ramp generator/signal converter, or the computer input circuit board). The failure does involve the output of the flow controller but is most likely the connector.
Failure to closely monitor flow controller output following previous flow controller replacement contributed to the failure to identify the fluctuations more promptly.
Equipment Reliability Issues also contributed to this failure. Subsequent to the replacement of the flow controller and connector, the power supply, the ramp generator/signal converter and the computer input circuit board, the condition has not reoccurred.
The root cause analysis associated with the RCIC flow controller performance is still under management review at the time of this report. A supplemental report will be issued, if required, if any additional causes and corrective actions are identified.
EVENT ANALYSIS
The RCIC system provides core cooling automatically or manually following Reactor Pressure Vessel (RPV) isolation. The RCIC system provides cooling for reactor pressures from 165 to 1215 pounds per square inch absolute (psia) (rated flow is 700 gallons per minute at 1118 psia). The RCIC system is designed to initiate and discharge, within 30 seconds, to provide the specified flow into the RPV at the specified pressure.
Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.3 requires that the system be operable, in Mode 1. With the RCIC system inoperable, the Required Action is to verify High Pressure Core Spray is Operable within an hour and to restore the RCIC system to operable status in 14 days or be in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Since the condition had been unrecognized from December 10, 2007, until January 14, 2008, these requirements were not completed within the required time.
The RCIC,system response is credited for several plant transients to provide RPV level control and decay heat removal until Residual Heat Removal systems are capable of providing decay heat removal. A conservative estimation of RCIC mission time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (maximum required operating duration based upon credited transient sequences and assumed in the Probabilistic Risk Assessment model).
The immediate investigation noted that the Flow Controller Output anomaly was repeating and the first occurrence was on December 21, 2007. The follow-up root cause investigation identified additional occurrences dating back to December 10, 2007. The anomaly has therefore been known to be present for 35 days. The investigation team 3 believes that the anomaly was most likely introduced into the system during the numerous flow controller changes that were performed to address RCIC flow controller tuning concerns during November and December 2007.
A bounding evaluation was performed assuming automatic operation of RCIC has been unavailable for 48 days (November 28, 2007, through January 14, 2008). Operation of the RCIC system using manual control was available and demonstrated following the plant shutdown in November. The probability of the RCIC manual control failure as a result of human error was included in the evaluation. The resulting Incremental Conditional Core Damage Probability (ICCDP) was determined to be 1.84E-08. The Incremental Conditional Large Early Release Probability (ICLERP) by definition can not be greater than the ICCDP.
Configurations with a core damage probability of less than 1.0E-06 and a large early release probability of less than 1.0E-07 are not considered to be risk significant events, therefore this event is considered to be of very low risk significance.
CORRECTIVE ACTIONS
The Bailey 701 flow controller and connecter, the power supply, the ramp generator/signal converter, and the computer input circuit board were replaced.
Plant operating staff is currently monitoring the flow controller output via the plant computer on a daily basis. This check has been included in the control room operator rounds.
The Bailey 701 Flow Controllers will be replaced with new controllers no later than the end of Refuel Wage 12, scheduled for the spring of 2009.
The RCIC Control System Tuning procedure will be revised to specify the use of the plant computer to monitor RCIC flow controller output following replacement or adjustment to a controller.
PREVIOUS SIMILAR EVENTS
Two LERs, Automatic Reactor Protection System Actuation Due to Feedwater Control Power Supply Failure, 2007-004-1 and Plant Startup With Inoperable Reactor Core Isolation Cooling System, 2007-005 document events in 2007. Both conditions were related to improper tuning of the RCIC Bailey 701 flow controller. While both conditions resulted in the flow controller being replaced, the purpose of replacement was to minimize the out of service time for the tuning and not the result of a component failure. For the January 14, 2008, event described in this LER the failure cause is considered a component failure therefore the cause is different than the previous tuning issues. Since the cause is different, the corrective actions from LER 2007-004-1 and LER 2007-005 would not have been expected to prevent this event.
Condition Report 06-00318 documents an event in which the RCIC pump did not respond properly when placed in Automatic. The cause of that event was determined to be dirty contacts on the balance resistor causing the Bailey flow controller to drift low. The issue in �NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER 3.) been due to failure of a flow controller component. The corrective actions from Condition Report 06-00318 include replacing the Bailey flow controller with a new flow controller from another manufacturer. This corrective action has not yet been completed and is being tracked for completion in Refuel Outage 12.
COMMITMENTS
There are no regulatory commitments contained in this letter. Any actions discussed in this document that represent intended or planned actions are described for the NRC's information, and are not regulatory commitments.
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05000413/LER-2008-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | 05000334/LER-2008-001 | Control Room Envelope Air Intake During Normal Operation Higher Than Assumed In Design Basis Accident Dose Calculations | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000315/LER-2008-001 | Unit 1 Manual Reactor Trip due to Main Turbine Vibrations | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000272/LER-2008-001 | Inadvertent Start of an Emergency Diesel Generator During Testing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000251/LER-2008-001 | Manual Reactor Trip due to High Level in the 4A Steam Generator | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000247/LER-2008-001 | Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused by Loss of Feedwater Flow as a Result of Feedwater Pump Speed Control Malfunction | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000530/LER-2008-001 | Manual Reactor Trip when Removing a Degraded CEDM MG Set from Service | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000457/LER-2008-001 | 2A Essential Service Water Train Inoperable due to Strainer Fouling from Bryozoa Deposition and Growth | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000454/LER-2008-001 | Technical Specification Non-Compliance of Containment Sump Monitor Due to Improper Installation During Oriainal construction | | 05000440/LER-2008-001 | ' Condition Prohibited by Technical Specifications Due to Unrecognized Reactor Core Isolation Cooling Inoperability | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000483/LER-2008-001 | Containment Cooler Inoperability | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000412/LER-2008-001 | Unplanned Actuation of the Auxiliary Feedwater System During Plant Startup | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000374/LER-2008-001 | High Pressure Core Spray System Declared Inoperable Due to Failed Room Ventilation Supply Fan | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000323/LER-2008-001 | Reactor Trip Due to Main Electrical Transformer Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000305/LER-2008-001 | Pressurizer PORV and Reactor Coolant System Vent Valves Appendix R Spurious Operation Concern | | 05000271/LER-2008-001 | Crane Travel Limit Stops not Installed as Required by Technical Specifications due to an Inadequate Procedure | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000269/LER-2008-001 | Gas Void Found in High Pressure Injection System Suction Piping | | 05000263/LER-2008-001 | | | 05000261/LER-2008-001 | Appendix R Pathway Impassable due to Lock Configuration | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor | 05000361/LER-2008-001 | Valid actuation of Emergency Feedwater system following Main Feedwater pump trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000335/LER-2008-001 | Unattended Ammunition Discovered Inside Protected Area | | 05000456/LER-2008-001 | Technical Specification Non-Compliance Due to Inadequate Design of Auxiliary Feedwater (AF) Tunnel Access Covers Causing AF Valves Within the Tunnel to be Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000423/LER-2008-001 | Postulated Fire Scenario Results in Unanalyzed Condition - Pressurizer Overfill | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000369/LER-2008-001 | Potential Failure of Containment Isolation Valves (CIV) to Remain Fully Closed and Ino erable loner than allowed bCTechnical S ecification 3.6.3. | | 05000220/LER-2008-002 | Manual Reactor Scram Due to Loss of Reactor Pressure Control | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000261/LER-2008-002 | Manual Reactor Trip due to High Turbine Vibrations | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000270/LER-2008-002 | Main Steam Relief Valves Exceeded Lift Setpoint Acceptance Band 050002 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000305/LER-2008-002 | 5 . Blocked Open Steam Exclusion Door Results in Postulated Inoperability of Safety Systems | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000336/LER-2008-002 | Unplanned LCO Entry - Three Charging Pumps Aligned for Injection With the Reactor Coolant System Temperature Less than 300 Degrees F. | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000354/LER-2008-002 | BLOWN lE INVERTER MAIN FUSE WITH ONE EMERGENCY DIESEL GENERATOR INOPERABLE CAUSES LOSS OF CONTROL ROOM EMERGENCY FILTRATION LOSS OF SAFETY FUNCTION | | 05000395/LER-2008-002 | Control Room Normal and Emergency Air Handling Systems Inoperable Due to Pressure Boundary Breach | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000382/LER-2008-002 | Inoperable Steam Generator Narrow Range Level Channels | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000369/LER-2008-002 | 'Energy® BRUCE H HAMILTON Vice President McGuire Nuclear Station Duke Energy Corporation MGOIVP 112700 Hagers Ferry Road Huntersville, NC 28078 704-875-5333 704-875-4809 fax bhhamilton@duke-energy.corn August 21, 2008
U.S. Nuclear Regulatory Commission
ATTENTION: Document Control Desk
Washington, D.C. 20555
Subject: Duke Energy Carolinas, LLC
McGuire Nuclear Station, Unit 1
Docket No. 50-369
Licensee Event Report 369/2008-02, Revision 0
Problem Investigation Process No.: M-08-03862
Pursuant to 10 CFR 50.73 Sections (a)(1) and (d), attached
is Licensee Event Report (LER) 369/2008-02, Revision 0,
regarding the Unit 1 Reactor trip on June 26, 2008 due to
the 1B Reactor Coolant Pump Motor trip which was caused by
a failed surge capacitor.
This report is being submitted in accordance with 10 CFR
50.73 (a)(2)(iv)(A). This event is considered to be of no
significance with respect to the health and safety of the
public. There are no regulatory commitments contained in
this LER.
If questions arise regarding this LER, contact Lee A. Hentz
at 704-875-4187.
Very truly yours,
Bruce H. Hamilton
Attachment
www. duke-energy. corn U.S. Nuclear Regulatory Commission
August 21, 2008
Page 2
cc: L. A. Reyes, Regional Administrator
U.S. Nuclear Regulatory Commission, Region II
Sam Nunn Atlanta Federal Center
61 Forsyth Street, SW, Suite 23T85
Atlanta, GA 30303
J. F. Stang, Jr. (Addressee Only)
Senior Project Manager (McGuire)
U.S. Nuclear Regulatory Commission
Mail Stop O-8G9A
Washington, DC 20555
J. B. Brady
Senior Resident Inspector
U.S. Nuclear Regulatory Commission
McGuire Nuclear Station
B. 0. Hall, Section Chief
Radiation Protection Section
1645 Mail Service Center
Raleigh, NC 27699
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVEDBYOMB:NO.3150-0104 EXPIRES: 08/31/2010 (9-2007) Estimated burden per response to comply with this mandatory collection request: 50 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send ' LICENSEE EVENT REPORT (LER) comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to infocollects@nrc.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 (See reverse for required number of means used to impose an information collection does not display a currently valid OMB control digits/characters for each block) number, the NRC may not conduct or sponsor, and a person is not required to respond to,' the information collection. 1, FACILITY NAME 2. DOCKET NUMBER I 3. PAGE 05000- ' 1 8McGuire Nuclear Station, Unit 1 _ 0369' OF .4, TITLE . . Unit 1 Reactor Trip due to the 1B Reactor Coolant Pump Motor Trip which was caused by a
failed Surge Capacitor | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000440/LER-2008-002 | Inoperable Emergency Closed Cooling System Results In Condition Prohibited By Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000454/LER-2008-002 | Technical Specification Non-Compliance Due to Inadequate Design of Auxiliary Feedwater (AF) Tunnel Access Covers Causing AF Valves Within the Tunnel to be Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000298/LER-2008-002 | Technical Specification Prohibited Condition Due to Safety Relief Valve Test Failure | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000237/LER-2008-002 | Unit 2 High Pressure Coolant Injection System Declared Inoperable | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000272/LER-2008-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000361/LER-2008-002 | Disturbance on the Pacific DC Intertie cause offsite power frequency to dip below operability limits | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000362/LER-2008-003 | Missed TS completion time results in TS Violation | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000220/LER-2008-003 | Power Supplies for Drywell Pressure Indication not Qualified for Required Post-Accident Operating Duration | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000529/LER-2008-003 | Technical Specification - Limiting Condition for Operation 3.0.3 for Greater Than 1 Hour | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000293/LER-2008-003 | | | 05000263/LER-2008-003 | | | 05000423/LER-2008-003 | Automatic Reactor Trip During Shutdown for Refueling Outage 3R12 ., | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000251/LER-2008-003 | . Class 1 Weld Leak Due to Fatigue and Completion of Technical Specification Required Shutdown | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000282/LER-2008-003 | Loss of AFW Safety Function and Condition Prohibited by Technical Specifications Due to Mispositioned Isolation Valve | 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 | 05000286/LER-2008-003 | Automatic Actuation of Emergency Diesel Generator 33 During Surveillance Testing Caused by .Inadvertent Actuation of the Undervoltage Sensing Circuit on 480 Volt AC Safeguards Bus 5A | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000361/LER-2008-003 | Disturbance on the Pacific DC Intertie cause offsite power frequency to dip below operability limits | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000305/LER-2008-003 | Door Bottom Seal Failure Results in Inoperability of Control Room Ventilation System | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat |
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