On October 17, 2010 at 0209 hours0.00242 days <br />0.0581 hours <br />3.455688e-4 weeks <br />7.95245e-5 months <br />, during startup following the completion of the End of Cycle 17 Refueling Outage, Unit 2 entered Mode 4 with the required motor driven train (manually-actuated) of the Auxiliary Feedwater ( AFW) System inoperable. The required train's (2A or 2B) pump discharge valve was not in the correct position (open) as required by plant Technical Specifications (TS) and plant procedures. The affected valves were
- discovered to be in the closed position at 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br />. Following the discovery of this event, the affected valves were repositioned to the correct poition. The root cause of this event was determined to be that the procedure for AFW System operation was not structured to provide clear guidance to the operators for the status of AFW System alignments for Mode 4 and Mode 3. The affected procedure will be revised to provide clear guidance for the determination of AFW System alignments for entering these modes. Throughout this event, the affected valves were capable of being manually positioned according to plant procedures in the unlikely event of a plant transient requiring a manual start of the AFW System in Mode 4. This event was limited to plant operation in Mode 4 only. The health and safety of the public were not adversely affected by this event'.
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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 •Docket Number |
Event date: |
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Report date: |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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4142010002R00 - NRC Website |
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BACKGROUND
This event is being reported under the folloWing criterion:
10 CFR 50.73(a)(2)(i)(B), any operation or condition which was prohibited by the plant's Technical Specifications (TS).
Catawba Nuclear Station Unit 2 is a Westinghouse four-loop Pressurized Water Reactor (PWR) [El IS: RCT].
The Auxiliary Feedwater (AFW) System [EllS: BA] (Duke Energy designation "CA") supplies feedwater to the steam generators [EDS: SG] to removerdecay heat from the Reactor Coolant System [EllS: AB] upon the loss of normal feedwater supply. The AFW pumps [EllS: P] take suction through suction lines from the Condensate Storage System (CSS) [EllS: KA] and pump to the steam generator secondary side. The normal supply of water to the AFW pumps is from the condensate system. The supply valves [EllS: V] are open with power removed from the valve operator. The assured source of water to the AFW System is supplied by the Nuclear Service Water System (NSWS) [EllS: BI]. The turbine and motor driven pump discharge lines to each individual steam generator join into single lines outside containment. These individual lines penetrate the containment and enter each steam generator through the auxiliary feedwater nozzle. The steam generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere from the steam'generators via the Main Steam Safety Valves (MSSVs) [EllS: SA] or the steam generator Power Operated Relief Valves (PORVs) [EllS: SA]. If the main condenser [EllS: COND] is available, steam may be released via the steam dump valves [EllS: JI] and recirculated to the hotwell.
The AFW System consists of two motor driven AFW pumps and one steam turbine driven pump configured into three trains. Each of the motor driven pumps supplies 100% of the flow requirements to two steam generators, although each pump has the capability to be realigned to feed other steam generators. The turbine driven pump provides 200% of the flow requirements and supplies water to all four steam generators. Travel stops are set on the steam generator flow control valves [EllS: FCV] such that the pumps can supply the minimum flow required without exceeding the maximum flow allowed. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system. Each motor driven AFW pump is powered from an independent Class 1 E power supply. The steam turbine driven AFW pump receives steam from two main steam lines upstream of the Main Steam Isolation Valves (MSIVs) [EllS: SB]. Each of the steam feed lines will supply 100% of the requirements of the turbine driven AFW pump.
The AFW System is capable of supplying feedwater to the steam generators during normal unit startup, shutdown, and hot standby conditions. One turbine driven pump at full flow is sufficient to remove decay heat and cool the unit to Residual Heat Removal (RHR) [EDS: BP] entry conditions. During unit cooldown, steam generator pressures and main steam pressures decrease simultaneously. Thus, the turbine driven AFW pump with a reduced steam supply pressure remains fully capable of providing flow to all steam generators. Thus, the requirement for diversity in motive power sources for the AFW System is met.
The AFW System is designed to supply sufficient water to the steam generators to remove decay heat with steam generator pressure at the lowest setpoint of the MSSVs plus 3% accumulation. Subsequently, the AFW System supplies sufficient water to cool the unit to RHR entry conditions, with steam released through the steam generator PORVs or MSSVs.
The motor driven AFW pumps actuate automatically on steam generator water level low-low in one out of four steam generators by the Engineered Safety Features Actuation System (ESFAS) [EllS: JE]. The motor driven pumps also actuate on loss of offsite power, safety injection, and trip of all Main Feedwater (MFW) [El IS: SJ] pumps. The turbine driven AFW pump actuates automatically on steam generator water level low-low in two out of four steam generators and on loss of offsite power.
TS 3.7.5 governs the AFW System. Limiting Condition for Operation (LCO) 3.7.5 requires three AFW trains to be operable in Modes 1, 2, and 3. In Mode 4 when the steam generator(s) are relied upon for heat removal, only one AFW train, which includes a motor driven pump, is required to be operable. Since the ESFAS instrumentation that actuates the AFW System is not required to be operable in Mode 4, manual actuation of the required AFW train in this mode is sufficient.
LCO 3.0.4.b allows mode changes into Modes 2, 3, and 4 with an inoperable AFW train; however, LCO 3.0.4.b is not applicable when entering Mode 1.
Condition E states that with the required AFW train inoperable in Mode 4, action must be immediately initiated to restore the AFW train to operable status.
Surveillance Requirement (SR) 3.7.5.1 requires a verification that each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. The SR is modified by a Note that states that it is not applicable to automatic valves when thermal power is The SR has a Frequency of 31 days.
Procedure OP/2/A/6250/002, "Auxiliary Feedwater System", governs the operation of the Unit 2 AFW System.
The procedure contains a number of enclosures, several of which are relevant to the event described in this LER.
These are Enclosure 4.1, "Placing the CA System in Standby Readiness", Enclosure 4.3, "Manual Operation of the Motor Driven Auxiliary Feedwater Pumps When Aligned For Standby Readiness", and Enclosure 4.5, "Manual Operation of the Motor Driven Auxiliary Feedwater Pumps When Not Aligned For Standby Readiness".
On October 17, 2010, when this event occurred, Unit 2 was in Mode 4 at 0% power operation, in the process of starting up following the completion of the End of Cycle (EOC) 17 Refueling Outage (RFO).
EVENT DESCRIPTION
Date/Time� Event 10/15/2010/0438� Operations began the pre-Mode 4 AFW valve verification for startup following completion of the EOC 17 RFO.
0530� The 2A and 2B motor driven AFW pump breakers were racked in.
1000� The 2A and 2B motor driven AFW pump discharge valves were opened per OP/2/A/6250/002, Enclosure 4.1.
1011�A log entry was made that the AFW System was in standby readiness.
1052�The pre-Mode 4 checklist was completed.
1900�The secondary execution Senior Reactor Operator (SRO) mistakenly concluded that the motor driven AFW pumps were not in standby readiness following a review of OP/2/N6250/002, Enclosure 4.1.
1930�The secondary execution SRO conducted a task preview of OP/2/A/6250/002, Enclosure 4.5 with a Nuclear Equipment Operator (NEO).
2103 �The balance of plant Reactor Operator (RO) ran the 2A and 2B motor driven AFW pumps.
2104�The balance of plant RO received concurrent verification for closing the 2A and 2B motor driven AFW pump discharge valves and closed the valves. The Regulatory Guide (RG) 1.47 bypass panel alarmed.
2200�The control room SRO approved the completion of OP/2/A/6250/002, Enclosure 4.5.
10/15 or 16/2010/* The pre-Mode 4 checklist RG 1.47 bypass panel step was signed off. (The AFW System conditions necessary for the signing of this step were not met at the actual time of Mode 4 entry.) 10/16/2010/0600�Shift turnover occurred. The closed valves were not identified.
1800�Shift turnover occurred. The closed valves were not identified.
Mode 4 checklist and unit startup procedural steps were signed off.
10/17/2010/0209�Unit 2 entered Mode 4.
0630�Shift turnover occurred. The closed valves were not identified.
1100�Operations began the pre-Mode 3 AFW valve verification.
1330�The 2A and 2B motor driven AFW pump discharge valves were discovered closed during preparation for AFW autostart alignment per OP/2/N6250/002, Enclosure 4.1 (Mode 3 alignment portion). This rendered both motor driven AFW trains inoperable (only one train was required to be operable for Mode 4, however).
1408�The 2A motor driven AFW train was restored to operable status by opening its pump discharge valve.
1414�The pre-Mode 3 AFW valve verification was completed.
CAUSAL FACTORS
The root cause of this event was determined to be that procedure OP/2/A/6250/002 was not structured to provide clear guidance to the operators for the status of AFW System alignments for Mode 4 and Mode 3. Specifically:
- Enclosure 4.1 did not include clear direction as to when the system is aligned for standby readiness for Mode 4 and that the remaining steps will align AFW for Mode 3.
- Enclosure 4.3 did not clarify that the enclosure can be used when Enclosure 4.1 is performed to the point of aligning AFW for Mode 4.
- The title of Enclosure 4.5 did. not have the word "Not" in all capitals, bold, and underscored as required by the Procedure Writer's Manual.
- Enclosure 4.5 did not include an initial condition to verify the system is not aligned for standby readiness per Enclosure 4.1.
Per TS 3.7.5, there are different requirements for standby readiness for Mode 4 as opposed to Mode 3. Mode 4 requires one motor driven AFW train aligned with manual start capability, whereas Mode 3 requires both motor driven AFW trains aligned with automatic start capability. Enclosure 4.1 contained the configuration for both Mode 4 standby readiness and for Mode 3 standby readiness without a clear delineation as to when the Mode 4 alignment was completed. Without this clear delineation, the selection of the correct enclosure for manual operation of the motor driven AFW pumps depended upon the knowledge of the individual.
There were several contributing causes which contributed to this event and several missed opportunities to prevent this event from occurring. These included failure of control room personnel to exercise appropriate human performance tools, incorrect conclusion regarding the status of the motor driven AFW pumps by the secondary execution SRO, failure of the balance of plant RO to complete an appropriate turnover on the night of the motor driven AFW pump runs, and failure of control room personnel to conduct an adequate review of the RG 1.47 bypass panel during shift turnover and immediately prior to changing modes.
CORRECTIVE ACTIONS
Immediate:
1.�The affected valves were opened to satisfy TS and procedural requirements.
Subsequent:
1. Information concerning this event, including lessons learned, was disseminated to affected personnel.
2. An Operations Guide was issued to provide Operations Shift Manager (OSM) oversight of activities associated with AFW System alignments and expectations for RG 1.47 bypass panel review.
Planned:
1. Procedure OP/2/N6250/002 (as well as the corresponding Unit 1 procedure) will be revised to provide clear guidance for the determination of AFW System alignments for entering Mode 4 and Mode 3. The revisions will address the bulleted items discussed above.
2. The pre-outage briefing and Operations training will be revised to include the importance of performing pre job briefs and post-job reviews, RG 1.47 bypass panel reviews with relation to plant status, and control room personnel verification and validation of work originated by other groups.
3. A procedure change will be made to ensure that a verification is performed immediately prior to entering Mode 4 that no annunciators on the RG 1.47 bypass panel indicate an inoperable system preventing Mode 4 entry.
There are no NRC commitments contained in this LER.
SAFETY ANALYSIS
This event had no safety significance. The Catawba TS require automatic actuation capability of the AFW System in Modes 1, 2, and 3 only. In Mode 4, manual actuation capability of one motor driven AFW train is all that is required. Throughout this event, the control room operators had the ability to manually start the motor driven AFW pumps had AFW initiation been necessary. The operators would have manually aligned the motor driven AFW pump discharge valves to establish flow to the steam generators had it been needed. The operators have the ability to open the discharge valves and the flow control valves from the control room. While unlikely, the following events could have theoretically occurred while in Mode 4, resulting in the need for AFW System operation:
- Reactor trip or inadvertent safety injection below the P-11 permissive
- Loss of coolant accident Manual operation of the motor driven AFW pump discharge valves in response to these events would have been accomplished in accordance with existing procedural guidance had it been required. During this event while the discharge valves were in the closed position, the valves were powered, there was no maintenance being performed on the valves, and there were no administrative impediments (e.g., tags) that would have prevented the operators from opening the valves at any time while in Mode 4. All other AFW System valves were aligned , properly to allow flow to the steam generators without additional operator intervention.
This event did not affect the health and safety of the public.
ADDITIONAL INFORMATION
Within the previous three years, there have been other LER events involving TS violations; however, the specific circumstances surrounding those events and the corrective actions taken in response to those events could not have prevented this event from occurring. This event is therefore considered to be non-recurring.
Energy Industry Identification System (EIIS) codes are identified in the text as [El IS: XX]. This event is not considered reportable to the Equipment Performance and Information Exchange (EPIX) program.
This event is not considered to constitute a Safety System Functional Failure. The affected AFW trains remained capable of manual actuation throughout this event. There was no release of radioactive material, radiation overexposure, or personnel injury associated with the event described in this LER.
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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 50.73(a)(2)(ii) | 05000374/LER-2010-001 | High Pressure Core Spray System Declared Inoperable Due to Failed Room Ventilation Control Relay | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000373/LER-2010-001 | Unauthorized Individual Gained Access to the Protected Area. | | 05000370/LER-2010-001 | Loose connection in a panel board serving a Solid State Protection System Train concurrent with redundant train maintenance could have prevented fulfillment of a safety function. | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000261/LER-2010-001 | Emergency Diesel Generator Inoperable in Excess of Technical Specifications Allowed Completion Time | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000263/LER-2010-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000255/LER-2010-001 | Potential Loss of Safety Function Due to a Service Water Pump Shaft Coupling Failure | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000266/LER-2010-001 | Engineered Safety Features Steam Line Pressure Dynamics Modules Discovered Outside of Technical Specification Values | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000306/LER-2010-001 | Unit 2 Turbine Trip during Reactor Shutdown Resulting in a Reactor Scram | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000305/LER-2010-001 | Safety Injection Pump Recirculation Line Isolation Results in Violation of Technical Specifications | | 05000298/LER-2010-001 | Cooper Nuclear Station 05000298 1 of 4 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000269/LER-2010-001 | Standby Shutdown Facility Letdown Line Orifice Strainer Blocked by Valve Gasket Material | 10 CFR 50.73(a)(2)(i)(b) | 05000282/LER-2010-001 | Unanalyzed Condition Due to Postulated High Energy Line Break On Cooling Water System | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000277/LER-2010-001 | Multiple Slow Control Rods Results in Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i) | 05000361/LER-2010-001 | Broken Manual Valve Prevents Timely Condensate Storage Tank Isolation | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000483/LER-2010-001 | Emergency Core Cooling System MODE 4 Operating Practices Prohibited by current Technical Specification 3.5.3 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000498/LER-2010-001 | Unit Shutdown Required by Technical Specifications | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000316/LER-2010-001 | Valid Actuation of Auxiliary Feedwater System in Response to Valid Steam Generator Low-Low Levels | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000321/LER-2010-001 | Corrosion Induced Bonding Results in Safety Relief Valve Lift Setpoint Drift | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000336/LER-2010-001 | Millstone Power Station Unit 2 Reactor Trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000413/LER-2010-001 | Technical Specification Violation Associated with Failure to Perform Offsite Circuit Verification | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000440/LER-2010-001 | Invalid Isolation Signal Results in Shutdown Cooling Interruption | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | 05000424/LER-2010-001 | Breaker Failure Results in I B Train Containment Cooling System Being Declared Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000416/LER-2010-001 | Automatic Reactor Scram On Decreasing Reactor Water Level Due To Inadvertent Reactor Feed Pump Trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000249/LER-2010-001 | OPRM Power Supply Failure during Maintenance Results in Unit 3 Automatic Reactor Scram | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000251/LER-2010-001 | Two Shutdown Bank Rods Were Dropped from Fully Withdrawn Position | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv), System Actuation | 05000261/LER-2010-002 | Plant Trip due to Electrical Fault | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000255/LER-2010-002 | Condition that Could Have Prevented the Fulfillment of a Safety Function | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000335/LER-2010-002 | Opened ECCS Boundary Door in Violation of Identified Compensatory Measures | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000334/LER-2010-002 | 270 Degree Circumferential Flaw Found on Residual Heat Removal System Drain Valve Socket Weld | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000316/LER-2010-002 | Containment Divider Barrier Seal Mounting Bolts Not Properly Installed | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000250/LER-2010-002 | Fuel Transfer Pump Failure Renders 3B Emergency Diesel Generator Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000269/LER-2010-002 | Manual Reactor Trip due to 1A1 and 1A2 Reactor Coolant PumDHigh Vibration Indication | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000315/LER-2010-002 | Manual Auxiliary Feedwater Actuation in Response to Main Feedpump Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000271/LER-2010-002 | Inoperability of Main Steam Safety Relief Valves due to Degraded Thread Seals | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000277/LER-2010-002 | Improperly Fastened Rod Hanger Results in Inoperable Subsystem of Emergency Service Water | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000413/LER-2010-002 | Discovery of Reactor Coolant System Pressure Boundary Leak at Thermowell 1NCTW5850 Seal Weld. | | 05000282/LER-2010-002 | Postulated Flooding of Unit 1 Fuel Oil Transfer Pump Motor Starters Could Have Resulted In Reduced Fuel Oil Inventory | | 05000414/LER-2010-002 | Duke Energy Corporation Catawba Nuclear Station 4800 Concord Road York, SC 29745 803-701-4251 803-701-3221 fax December 15, 2010 U.S. Nuclear Regulatory Commission
Attention: Document Control Desk
Washington, D.C. 20555
Subject:�Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Unit 2
Docket No. 50-414
Licensee Event Report 414/2010-002
Pursuant to 10 CFR 50.73(a)(1) and (d), attached is Licensee Event Report 414/2010-002, Revision 0 entitled, "Technical Specification Violation Involving Mode Change with Inoperable Auxiliary Feedwater System Train Due to Closed Pump Discharge Valves". This report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B). There are no regulatory commitments contained in this letter or its attachment. This event is considered to be of no significance with respect to the health and safety of the public. If there are any questions on this report, please contact L.J. Rudy at (803) 701-3084. Sincerely, faius4- A James R. Morris LJR/s Attachment www.duke-energy.corn (14 Document Control Desk Page 2 December 15, 2010 xc (with attachment): L.A. Reyes Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 J.H. Thompson (addressee only) NRC Project Manager U.S. Nuclear Regulatory Commission Mail Stop 8-G9A 11555 Rockville Pike Rockville, MD 20852-2738 G.A. Hutto, Ill NRC Senior Resident Inspector Catawba Nuclear Station INPO Records Center 700 Galleria Place Atlanta, GA 30339-5957 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010) Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send commentsLICENSEE EVENT REPORT (LER) regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mail to infocollectssesource@nre.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used(See reverse for required number of to impose an information collection does not display a currently valid OMB control number, the NRCdigits/characters for each block) may not conduct or sponsor, and a person is not required to respond to, the info(mation collection. 1.. FACILITY NAME 2. DOCKET NUMBER I3. PAGE Catawba Nuclear Station, Unit 2 05000414 1 OF 7 4. TITLE Technical Specification Violation Involving Mode Change with Inoperable Auxiliary Feedwater System Train Due to Closed Pump Discharge ValvesD • | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000306/LER-2010-002 | Unit 2 Turbine Shutdown Due To the Loss of a Main Feed Water Pump That Resulted in a Reactor Scram | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000440/LER-2010-002 | Piping Leak Results in Condition Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000382/LER-2010-002 | Main Feedwater Isolation Valve B exceeded allowed outage time due to tubing connection failure | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000370/LER-2010-002 | ref Energy® REGIS T. REPKO Vice President McGuire Nuclear Station Duke Energy MGO1VP / 12700 Hagers Ferry Rd. Huntersville, NC 28078 980-875-4111 980-875-4809 fax regis.repko(Codu ke-energy.corn 10 CFR 50.73 May 10, 2011 U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, D.C. 20555 Subject: D Duke Energy Carolinas, LLC McGuire Nuclear Station, Unit 2 Docket Nos. 50-370 Licensee Event Report (LER) 370/2010-02, Supplement 1 Problem Investigation Process (PIP) M-10-05982 Pursuant to 10 CFR 50.73 Sections (a) (1) and (d), attached is Supplement 1 to Licensee Event Report 370/2010-02, regarding past inoperability of the Unit 2 "A" Train Nuclear Service Water System and satisfies the commitment to supplement the LER following completion of the root cause analysis This supplement to LER 370/2010-02 supersedes the LER previously submitted December 20, 2010. Completion of the root cause analysis has not affected the original reporting criteria which was completed in accordance with 10 CFR 50.73 (a) (2) (i) (B), an Operation Prohibited by Technical Specifications, and 10 CFR 50.73 (a) (2) (v) (B), any Event or Condition That Could Have Prevented Fulfillment of the Safety Function needed to remove residual heat. Additionally, the supplement did not affect the significance of the event which was considered to be of no significance with respect to the health and safety of the public. There are no regulatory commitments contained in this report. If questions arise regarding this LER, contact Rick Abbott at 980-875-4685. Very truly yours, Zi1:77 Regis T. Repko Attachment www. duke-energy. corn U.S. Nuclear Regulatory Commission May 10, 2011 Page 2 cc:�V. M. McCree, Regional Administrator U.S. Nuclear Regulatory Commission, Region II
Marquis One Tower
245 Peachtree Center Ave., NC, Suite 1200
Atlanta, Georgia 30303-1257
Jon H. Thompson (Addressee Only)
Senior Project Manager (McGuire)
U.S. Nuclear Regulatory Commission
11555 Rockville Pike
Rockville, MD 20852-2738
J. B. Brady
Senior Resident Inspector
U.S. Nuclear Regulatory Commission
McGuire Nuclear Station
W. L. Cox Ill, Section Chief North Carolina Department of Environment and Natural Resources Division of Environmental Health Radiation Protection Section 1645 Mail Service Center Raleigh, NC 27699-1645 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB. NO 3150-0104 EXPIRES: 08/31/2013 (10-2010) Estimated burden per response to comply with this mandatory collection request: SO hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the FOIA/Privacy Section (T-5 F53), U.S. Nuclear Regulatory Commission. Washington, DC 20555-0001, or by Internet e-mail to info (See reverse for required number of collects resmirceOnrc.gov, and to the Desk Officer, Office of Information and Regulatory digits/characters for each block) Affairs, NEOB-10202, (3150-01041, Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. LICENSEE EVENT REPORT (LER) 1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE McGuire Nuclear Station,2Unit 2 05000-212
0370 OF-7 4. TITLE Unit 2 Nuclear Service Water System "A" Train Past Inoperable due to
Failed Strainer Differential Pressure Instrument. | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000369/LER-2010-002 | | | 05000456/LER-2010-002 | Limiting Condition for Operation Action Not Completed Within the Required Time | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000249/LER-2010-003 | Steam Leak Results in HPCI Inoperability | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000251/LER-2010-003 | Damaged Speed Sensor Caused the 4A Emergency Diesel Generator to be Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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