On 7/02/09, the Nuclear System Protection System (NSPS) Self-Test System identified a failure of the Division 3 High Pressure Core Spray ( HPCS) logic circuit card (HPCS-1). This failure resulted in the output initiation signals from this card being blocked, preventing the safety function of automatic HPCS initiation and automatic start of the Division 3 emergency diesel generator ( DG) and the Division 3 Shutdown Service Water System pump. Operators declared the HPCS, the DG, and the SX systems inoperable but available and entered the applicable Technical Specification action requirements. The Reactor Core Isolation System was verified to be operable as required by Technical Specifications. The cause of the Division 3 logic circuit card failure was a knit-line delamination and associated cracks on a ceramic capacitor in the Power On Initialization circuit due to a manufacturing anomaly that limited the expected lifetime of this capacitor. The defective ceramic capacitor caused the output initiation signals from this card to be blocked, preventing automatic HPCS initiation signals. Corrective actions for this event include repairing the Division 3 logic circuit card that failed, obtaining a spare Division 3 logic circuit card and creating a Performance Centered Maintenance template and evaluating strategies for performing preventive maintenance activities on NSPS circuit cards. |
LER-2009-001, Safety Function Lost Due to Capacitor Failure on Circuit CardDocket Number |
Event date: |
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Reporting criterion: |
10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
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4612009001R00 - NRC Website |
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PLANT OPERATING CONDITIONS
Unit: 1 Event Dates: 7/2/09 Event Time: 7/2/09, 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Central Daylight Time Mode: 1 (Power Operation) Reactor Power: 97 percent
DESCRIPTION OF EVENT
At 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on 7/2/09, control room operators received an alarm [ALM] indicating the Nuclear Systems Protection System (NSPS) [JE] Self Test System (STS) failed. Initial investigation identified the reason for the failure was a failure of Division 3 logic circuit card 1PAP663BA16A114: High Pressure Core Spray (HPCS) [BG] Error Code 1 (HPCS-1) system test. The operators attempted to restart the STS and it failed again for the same reason. In accordance with station procedures, the Instrumentation Maintenance Department was required to investigate the second failure, and this task was assigned to the day-shift.
Issue Report 938015 was initiated to track investigation of this issue.
The NSPS consists of four independent and redundant divisions of safety-related solid-state circuitry used to scram the reactor and to initiate emergency core cooling systems and engineered safety feature systems. The STS is a testing and surveillance system designed to automatically and continuously monitor the NSPS functional circuitry. The STS provides the means to continuously monitor the logic circuit integrity and the circuit continuity of the NSPS systems once every 40 minutes.
At about 1035 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.938175e-4 months <br />, initial troubleshooting at the station determined the Division 3 STS logic was locked up and the Division 3 HPCS logic circuit card was identified to be the problem.
At 1108 hours0.0128 days <br />0.308 hours <br />0.00183 weeks <br />4.21594e-4 months <br />, the Operations shift manager held a preemptive discussion with the Main Control Room team concerning a manual start of HPCS with logic inoperable due to concerns with the Division 3 logic circuit card.
At 1137 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.326285e-4 months <br />, preparations commenced to allow on-site testing of the Division 3 logic circuit card and to develop a strategy to obtain vendor repair of the card if needed.
At 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />, a dedicated main control room operator was designated to manually start and initiate HPCS if required.
At 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br />, troubleshooting concluded that the most likely cause of the STS failure was malfunction of the Power On Initialization (P01) circuitry on the Division 3 logic circuit card. This failure resulted in the output initiation signals from this card being blocked, preventing automatic HPCS initiation and automatic start of the Division 3 emergency diesel generator [EK] [DG] and the Division 3 Shutdown Service Water System (SX) [BI] pump [P]. At this time, operators declared the HPCS, Division 3 DG, and Division 3 SX systems inoperable but available and entered the applicable Technical Specification action requirements, requiring restoration of HPCS within 14 days. The failure of the Division 3 logic circuit card does not prevent manual start of the HPCS pump, opening of the injection valve, manual start of the Division 3 DG or manual start of Division 3 SX pump using hand switches in the Main Control Room. Operators verified the Reactor Core Isolation System (RCIC) [BN] was operable as required by Technical Specification required actions.
_ At about 1608 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.11844e-4 months <br />, the Division 3 logic circuit card was removed for further site testing (using a GENRAD tester). During this site testing, the POI circuitry of the Division 3 HPCS logic circuit card failed the test.
Subsequently, the Division 3 logic circuit card was sent to the supplier for further troubleshooting and diagnostic testing. Supplier investigation of the Division 3 logic circuit card confirmed the card failure was due to degraded coupling ceramic capacitors [CAP] in the POI circuitry. The card was repaired by the supplier and returned to the station.
At 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br /> on 7/7/09, following completion of installation of the card, post-maintenance testing and restoration of the automatic functions of the card, HPCS, and Division 3 DG and SX were declared operable.
No other inoperable equipment or components directly affected this event.
CAUSE OF EVENT
The cause of the Division 3 logic circuit card failure is attributed to a knit-line delamination and associated cracks that caused a low insulation resistance of the ceramic capacitor in the POI circuit of the card that was installed in 1987. The ceramic capacitor is one of two coupling capacitors in series with a resistor to form a time delay that prevents erroneous signals from actuating field devices during circuit card power-up.
The purpose of the POI circuit is to set latches to a predetermined state during power-up or during a circuit card removal, and to inhibit the outputs to the field instrumentation during power-up to allow logic setting time. The circuit analysis determined that one degraded ceramic capacitor on the POI circuit caused the HPCS-1 circuit card to malfunction, blocking initiation signals for HPCS, Division 3 DG and Division 3 SX systems. The vendor failure analysis of the capacitor was unable to identify the cause of the knit-line delamination and cracks; however, further industry research shows the most probable cause of the defective capacitor is a manufacturing anomaly that limited the expected life of the capacitor.
SAFETY ANALYSIS
This event is reportable under the provisions of 10 CFR 50.73(a)(2)(v)(D) due to a condition that could have prevented fulfillment of the HPCS safety function to mitigate the consequences of an accident.
The HPCS and Division 3 DG and Division 3 SX systems were inoperable but available from 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on 7/2/09 until declaration of operable status at 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br /> on 7/7/09, except for a period of 71 minutes from 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> to 1311 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.988355e-4 months <br /> on 7/7/09. During this time, the HPCS and Division 3 DG and Division 3 SX systems were inoperable and unavailable while control power fuses for the HPCS pump were removed for installation of the repaired Division 3 logic circuit card.
During the times the HPCS and Division 3 DG and Division 3 SX systems were available, the manual start capability of the HPCS pump and the functions for opening the injection valve using hand switches in the main control room continued to be available for operator manual initiation if required. Additionally, the manual initiation functions for the Division 3 DG and SX systems were available in the main control room if required during this time.
The RCIC system was operable during this event. Although no credit is taken in the safety analysis for the RCIC System, it performs a similar function as HPCS but has reduced makeup capability. Nevertheless, it will maintain inventory and cool the core, while the Reactor Coolant System is still pressurized, following a reactor pressure vessel isolation. If HPCS fails to maintain reactor water level above Level 1, it is backed up by automatic initiation � of Automatic Depressurization System in combination with Low Pressure Coolant Injection [BO] and Low Pressure Core Spray [BM] systems; these systems remained operable during this event.
CORRECTIVE ACTION
The Division 3 logic circuit card that failed has been repaired and a spare Division 3 logic circuit card will be obtained.
A Performance Centered Maintenance template will be created and strategies for performing preventive maintenance activities will be evaluated for NSPS circuit cards.
PREVIOUS OCCURRENCES
The 7/2/09 event was a repeat failure of STS identified on 6/24/09 (in issue report 934532) that caused the STS system to stop testing. The STS 6/24/09 failure was an intermittent failure on the HPCS-1 circuit card that was reset and ran successfully in fully automatic test mode. This event was not reportable under the provisions of 10 CFR 50.73.
COMPONENT FAILURE DATA
Circuit Card Manufacturer: General Electric Nomenclature: HPCS-1 circuit card Manufacture Model Number: 147D8500G001 Part Number: 1PAP663BA16A114 Ceramic Capacitor Manufacturer: KEMET Electronic Corporation Nomenclature: Multi-Layer Ceramic Capacitor, one micro-Farad Manufacturer Model Number: 1CK06BX105K
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05000410/LER-2009-001 | Momentary Loss of Control Power to High Pressure Core Spray, Pump Due to Degraded Fuse Block Connection | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000266/LER-2009-001 | Component Coolina Water PumD Inoperable In Excess of Technical Specification Allowed Completion Time | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000263/LER-2009-001 | Containment Overpressure Not Ensured in the Appendix R Analysis | 10 CFR 50.73(a)(2)(ii) | 05000250/LER-2009-001 | Procedure Inadequacy Causes Control Room Ventilation Isolation Technical Specification Noncompliance | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000237/LER-2009-001 | Common Mode Failure of Reactor Building Isolation Dampers | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000530/LER-2009-001 | Manual Reactor Trip Due to a Loss of Instrument Air to the Containment Building | | 05000457/LER-2009-001 | Reactor Trip on Over Temperature Delta Temperature due to a Signal Spike on One Channel With Another Channel Placed in the Tripped Condition for Surveillance Testing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000413/LER-2009-001 | Both Trains of Chemical and Volume Control, Auxiliary Feedwater and Containment Spray Systems were Inoperable due to a Component Failure | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000412/LER-2009-001 | Equipment Operability for Steam Generator Tube Rupture Safety Analysis Not Met | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vi) | 05000461/LER-2009-001 | Safety Function Lost Due to Capacitor Failure on Circuit Card | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000382/LER-2009-001 | Waterford 3 Steam Electric Station 05000382 1 OF 3 | | 05000370/LER-2009-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000346/LER-2009-001 | Containment Air Cooler Fans Inoperable Due to Misapplication of Potter and Brumfield Rotary Relays | 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 | 05000336/LER-2009-001 | Reactor Trip Due to High Pressurizer Pressure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000305/LER-2009-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000281/LER-2009-001 | Manual Reactor Trip Initiated to Replace a Rod Control Data Logging Card | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000261/LER-2009-001 | Emergency Diesel Generator Inoperable in Excess of Technical Specifications Allowed Completion Time | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000220/LER-2009-001 | Failure to Implement Required Technical Specification Actions Associated with Failed Surveillance Test | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000389/LER-2009-001 | Unit 2 Main Feedwater Isolation Valves Stroke Time Potentially Affected by Temperature | 10 CFR 50.73(a)(2)(I)(B) | 05000321/LER-2009-001 | Pump Suction Swap for HPCI and RCIC Non-Conservative With Respect To Technical Specification Requirements | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000334/LER-2009-001 | Surveillance Test Inadvertently Violates Technical Specification 3.6.1 for Containment Operability | 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-2009-001 | III Duke Energy® BRUCE H HAMILTON Vice President McGuire Nuclear Station Duke Energy Corporation MGO1VP / 12700 Hagers Ferry Road Huntersville, NC 28078 704-875-5333 704-875-4809 fax bhhamilton@duke-energy.com June 24, 2009
U.S. Nuclear Regulatory Commission
ATTENTION: Document Control Desk
Washington, D.C. 20555
Subject: Duke Energy Carolinas, LLC
McGuire Nuclear Station, Units 1 and 2
Docket Nos. 50-369, 50-370
Licensee Event Report 369/2009-01, Revision 0
Problem Investigation Process (PIP) M-09-02216
Pursuant to 10 CFR 50.73 Sections (a) (1) and (d), attached
is Licensee Event Report 369/2009-01, Revision 0, regarding
the past inoperability of the Nuclear Service Water System
"A" Trains due to potential for strainer fouling.
This report is being submitted 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.
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 Rick Abbott
at 980-875-4685.
Very truly yours,
Bruce H. Hamilton
Attachment
www.duke-energy.corn m U.S. Nuclear Regulatory Commission
Date
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. H. Thompson, Jr. (Addressee Only)
Senior Project Manager (McGuire)
U.S. Nuclear Regulatory Commission
Mail Stop 0-8G9A
Washington, DC 20555
J. B. Brady
Senior Resident Inspector
U.S. Nucle'ar Regulatory Commission-
McGuire Nuclear Station
B. 0. Hall, Section Chief
Radiation Protection Section
1645 Mall Service Center.
Raleigh, NC 27699
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104t EXPIRES: 08/31/2010
(9-2007) Estimated burden per response to comply with this mandatory collection request: 50 hours. Repoded
lessons learned are incorporated into the licensing process and fed back to industry. Send 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 means used to impose an information (See reverse for required number of collection does not display a currently valid OMB control number, the NRC may not conduct or digits/characters for each block) sponsor, and a person is not required to respond to, the information collection. LICENSEE EVENT REPORT (LER) 1. FACILITY NAME 2. DOCKET NUMBER I 3. PAGE _McGuire Nuclear Station, . 0369 8 Unit 1 05000- OF 4. TITLE Nuclear Service Water System (NSWS)d
"A" Trains Past Inoperable when aligned
to the Standby Nuclear Service Water Pond due to'corrosion.
(SNSWP) | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | 05000456/LER-2009-001 | Steam Generator Tube Exceeding Plugging Criteria Remained In Service During Previous Cycle | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000395/LER-2009-001 | Inadequate Procedure Results In EDG Not Obtaining Maximum Load Required By Technical Specification | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000336/LER-2009-002 | Reactor Coolant System Pressure Boundary Leakage | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000373/LER-2009-002 | Loss of Shutdown Cooling Due to Spurious Closure of the Shutdown Cooling Suction Isolation Valve | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | 05000382/LER-2009-002 | Waterford 3 Steam Electric Station 05000382 10OF 4 | | 05000278/LER-2009-002 | Inoperable 'A' Wide Range Neutron Monitor Results in Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000287/LER-2009-002 | Unit 3 Trip Due to Generator Phase Differential Lockout | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000263/LER-2009-002 | | | 05000412/LER-2009-002 | Unacceptable Indications Identified During Reactor Vessel Head Inspection | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000354/LER-2009-002 | As Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000346/LER-2009-002 | Vibration Induced Failure of Temperature Instrument Results in Operation above Licensed Power Limit | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000261/LER-2009-002 | Failure to Complete Technical Specifications Required Action Within the Allowed Completion Time | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000334/LER-2009-002 | Feedwater Isolation Initiates Auxiliary Feedwater System During Refueling Shutdown | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000254/LER-2009-002 | Quad Cities Nuclear Power Station Unit 1 05000254 1 OF 5 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000305/LER-2009-002 | Steam Exclusion Door Blocked Open During Maintenance Activities | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000250/LER-2009-002 | Turkey Point Unit 3 05000250 1 of 10 | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 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 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000220/LER-2009-002 | High Pressure Coolant Injection System Initiation Following a Manual Turbine Trip Due to High Turbine Bearing Vibrations | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000298/LER-2009-002 | Manual Scram On Low Water Level Caused By Turbine Trip From Hydraulic Fluid Leak | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000305/LER-2009-003 | Containment Spray Pump A Inoperable At Degraded Voltage Protection Setpoint | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000395/LER-2009-003 | ..Potential Loss of Residual Heat Removal System Safety Function In Mode 4 Due To An Unanalyzed Condition0 | 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 | 05000389/LER-2009-003 | RCP 2B2 Lower Seal Cavity Line Leak | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000323/LER-2009-003 | Containment Sump Recirculation Valve Position Interlock Failure Due to Inadequate Testing | | 05000263/LER-2009-003 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000261/LER-2009-003 | Manual Reactor Trip Due to Failure of 'A' Steam Generator Level Module | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000440/LER-2009-003 | Reactor Recirculation Pump Failure Results in Manual Reactor Protection System Actuation | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000361/LER-2009-003 | Pressurizer Auxiliary Spray Failed Inservice Test | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | 05000457/LER-2009-003 | Drain Procedure for ECCS Suction Line Creates an Unanalyzed Condition Due to Inadequate Configuration Requirements | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000237/LER-2009-003 | Emergency Diesel Generator Oil Leak | 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 |
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