| 7-9-2010 | On May 10, 2010, during a normal start for local parallel operation of the 4A Emergency Diesel Generator ( EDG) for a test, frequency unexpectedly increased to approximately 63 Hz. Consequently, a normal shutdown of the 4A EDG was initiated. While shutting down, the 4A EDG locked out due to "Piston Cooling Oil Low Pressure." The apparent cause of the 4A EDG's uncontrolled speed/frequency increase and subsequent lock-out is a failure of the speed sensing magnetic pickup due to being set too close to the engine flywheel. A change to the maintenance procedure will be made to ensure that the run out of the engine flywheel gear teeth is checked and that the magnetic pickups are set with adequate clearance from the top of the highest gear tooth. The other three EDGs were assessed and determined not to be affected. Since the magnetic pickup was set during a maintenance activity, the 4A EDG was considered inoperable from the start of maintenance until the magnetic pickup was repaired. This resulted in a period of inoperability of approximately 16.9 days. The risk impact associated with the approximate 16.9 days the 4A EDG was considered inoperable results in an ICCDP (Incremental Conditional Core Damage Probability) of 5.6E-08 and an ICLERP (Incremental Conditional Large Early Release Probability) of 9.8E-11, well below NRC thresholds. Therefore, safety significance is considered to be very low. |
|---|
LER-2010-003, Damaged Speed Sensor Caused the 4A Emergency Diesel Generator to be Inoperable| Docket Number |
| Event date: |
5-10-2010 |
|---|
| Report date: |
7-9-2010 |
|---|
| Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
|---|
| 2512010003R00 - NRC Website |
|
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
DESCRIPTION OF THE EVENT
On May 10, 2010, during a normal start for local parallel operation of the 4A Emergency Diesel Generator (EDG) [EK, DG] for a test, frequency unexpectedly increased to approximately 63 Hz. Consequently, a normal shutdown of the 4A EDG was initiated. While shutting down, the 4A EDG locked out due to "Piston Cooling Oil Low Pressure.
The EDG operating procedure requires generator voltage to be adjusted slightly higher than bus [EB, BU] voltage. After performing this step, 4A EDG frequency increased to approximately 62 Hz. After unsuccessful attempts to decrease the speed of the 4A EDG by manipulating the governor [EK, DG, 65] control switch [EK, DG, HS], frequency had increased to approximately 63 Hz, and speed indicated approximately 930 rpm and appeared to be slowly rising. As a result, Operations personnel decided to shutdown the 4A EDG. While shutting down, the 4A EDG locked out. A review of the alarms indicated that the lock out was caused by "Piston Cooling Oil Low Pressure".
Condition Report (CR) 2010-12306 was initiated in response to the event.
This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B).
CAUSE OF THE EVENT
The apparent cause of the 4A EDG's uncontrolled speed/frequency increase and subsequent lock-out during the normal shutdown initiation is a failure of the speed sensing magnetic pickup [EK, DG, SIC] due to being set too close to the engine flywheel.
ANALYSIS OF THE EVENT
Background
Each Turkey Point unit has two associated EDGs. Technical Specification (TS) Limiting Condition for Operation 3.8.1.1.b requires a unit's two EDGs and one of the opposite unit's EDGs to be operable to provide standby electrical power for required equipment in support of plant operation in Modes 1-4. The safety related function of the EDGs is to automatically start and provide power to required safety related loads during a loss of offsite power in order to achieve and maintain safe shutdown of the reactor [AC, RCT].
Analysis Initial identification of potential causes focused on the 2301A Electronic Governor (EG) and Governor Motor Operated Potentiometer used for speed control of the 4A EDG. Instrumentation was specified and FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 4 05000251 2010 - 003 - installed to gather data on the function of those components during a loaded run. Additionally, the resistance of the 4A EDG magnetic pickup for the EG at the engine flywheel was measured. This measurement revealed that the subject magnetic pickup was OPEN (not within the expected range of 100 300 Ohms) and needed to be replaced. This particular magnetic pickup (MP2) provides a feedback input to the 2301A EG. A malfunction or failure of this magnetic pickup will impact EG function and EDG speed.
The 2301A EG is reverse acting and controls the hydraulic actuator [EK, DG, HCU] on the engine. The 2301A EG is connected to a 125VDC external power source, a magnetic pickup at the engine flywheel for a speed sensing source, and current and potential transformers for measuring the load carried by the generator.
When a magnetic material (in this case the engine flywheel gear teeth driven by the prime mover or crank shaft) passes through the magnetic field at the end of the magnetic pickup, a voltage is developed. The frequency of this voltage is translated by the speed control into a signal which accurately depicts the speed of the prime mover.
The magnetic speed sensor was inspected and found to have 0.007 inches from the tip of the probe completely worn off as if the pick-up had contacted the rotating flywheel gear. Combining the observed sensor damage, sensor clearance, flywheel run out and bearing clearance effects, it was determined that the magnetic probe contacted the flywheel, leading to the observed damage and erratic performance of the pickup.
A pressure switch [EK, DG, PS] provides the input to trip the 4A EDG in the event of low piston cooling oil pressure. This trip is enabled by the 90 rpm speed switch and is set at 12 psig falling. The piston cooling oil pump [EK, DG, P] is a positive displacement gear pump driven off of a common shaft with the main oil pump. Flow is directly a function of engine speed. The pump pressure is not regulated by any valves and is controlled by the resistance of the system and the viscosity of the oil. The piston cooling oil pump develops approximately 72-74 psig at 900 rpm and 174°F oil temperature.
The loss of the magnetic pickup sensor (MP2) signal was the direct cause of the event. It was functioning above minimum requirements at the beginning of the run on May 10, 2010 as the 4A EDG initially went to both idle and rated speed as demanded. However, as damage continued, the output of the speed sensor became erratic, driving the EG to respond to non-existent speed deficiencies to the point that engine speed increased beyond the swap over to the hydraulic governor and eliminated operational control of engine speed via the EG. However, the speed sensor was still functioning in some degraded state even at that time as the EG resumed immediate control upon demand for a normal shutdown. Only the EG and the shutdown solenoid can reduce engine speed after the diesel is on the hydraulic governor. There was no indication that the shutdown solenoid was energized; therefore, the EG likely resumed control. However, during the reduction to idle and cool down, engine speed continued to decay until a lockout was generated on low piston cooling oil pressure. Further degradation in speed sensor pickup is expected at lower speed and aggravated by accumulated damage. The lockout was considered valid and an indication that engine speed fell considerably below the 450 rpm target which would cause the piston cooling oil pump pressure to fall below its 12 psig trip setpoint. Following replacement of the speed sensor, all data gathered indicated normal, expected performance from each of the EDG control components.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) The magnetic pickups are set during each major maintenance outage (CMM) and, as was seen in this instance with the 4A EDG, an incorrect setting is self-revealing after a few hours of runtime. The last CMM performed on the 4B EDG was in January 2009. The last CMMs for the 3A and 3B EDGs were back-to back in September/October 2008. No similar issues concerning speed control have been encountered with the other three EDGs; therefore, there is reasonable assurance that this issue is limited to the 4A EDG.
Past Operability Assessment The 4A EDG was removed from service at approximately 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br /> on April 27, 2010 for maintenance.
During maintenance, the magnetic pickup was assumed to be dimensionally set within the procedural acceptance criteria, yet this was close enough to the engine flywheel gear teeth to make contact with the raised edges on the flywheel.
Had operation of the 4A EDG been demanded, there is no reasonable assurance that the magnetic pickup would have operated properly. As such, the diesel was never truly operable following maintenance. Since the failure mechanism was considered to have existed from the maintenance activity until return to service on May 14, 2010 at approximately 1043, this period exceeds the 14-day allowed outage time of TS 3.8.1.1, Action b and is a total time of 16.9 days (406.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />).
Reportability A review of the reporting requirements of 10 CFR 50.72 and 10 CFR 50.73 and NRC guidance provided in NUREG-1022, Revision 2, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73, was performed for the subject condition. As a result of this review, the condition is reportable as described below.
The 4A EDG is considered inoperable from April 27, 2010 at approximately 1230 to May 14, 2010 at approximately 1043. The condition placed Unit 4 in TS 3.8.1.1, Action b which requires restoration of "...the inoperable diesel generator to OPERABLE status within 14 days** or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />." This action was not met since the 4A EDG was inoperable for approximately 16.9 days.
10 CFR 50.73(a)(2)(i)(B) requires the reporting of :
"Any operation or condition which was prohibited by the plant's Technical Specifications except when:
(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or (3) The. Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (31 As none of the three exceptions to 10 CFR 50.73(a)(2)(i)(B) apply in this case, exceeding the allowed outage time of TS 3.8.1.1, Action b is reportable.
ANALYSIS OF SAFETY SIGNIFICANCE
The approximate 16.9 days the 4A EDG is considered inoperable has minimal safety significance. Other required equipment was determined to be operable during this period.
Probabilistic risk assessment yields a delta CDF (Core Damage Frequency) of 6.1E-07 per year that falls within Region III of Figure 3 of U.S. NRC Regulatory Guide 1.174 (RG 1.174), An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, November 2002, defined as representing a "very small change". This figure is meant to be applied to permanent changes to the plant. The 4A EDG being out of service for 406.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was a one time, temporary condition, making this delta CDF even less of a concern. The calculated value for ICCDP (Incremental Conditional Core Damage Probability) of 2.8E-08 is well below the ICCDP threshold specified by the NRC in U.S. NRC Regulatory Guide 1.177 (RG 1.177), An Approach for Plant-Specific, Risk- quantitative impact" for a permanent TS change. Again, the 4A EDG being out of service for 406.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was a one-time, temporary condition, making this ICCDP even less of a concern.
The delta LERF (Large Early Release Frequency) of 1.1E-09 per year falls within Region III of Figure 4 of RG 1.174, defined as representing a "very small change". This figure is meant to be applied to permanent changes to the plant. The 4A EDG being out of service for 406.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was a one-time, temporary condition, making this delta LERF even less of a concern. The calculated value for ICLERP (Incremental Conditional Large Early Release Probability) of 4.9E-11 is well below the ICLERP threshold specified by Again, the 4A EDG being out of service for 406.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was a one-time, temporary condition, making this ICLERP even less of a concern.
To approximate the effect of external events, the delta CDF and delta LERF were doubled, resulting in a delta CDF of 1.2E-06 per year, a delta LERF of 2.2E-09 per year, an ICCDP of 5.6E-08, and an ICLERP of 9.8E-11. This assumption caused the delta CDF to move into Region II of Figure 4 of RG 1.174, defined as representing a "small change", with the delta LERF, ICCDP, and ICLERP unchanged with respect to the significance thresholds. Given the fact that this figure is meant to be applied to permanent changes to the plant, and that the 4A EDG being out of service for 406.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was a one-time, temporary condition, the conclusions regarding minimal impact were unchanged.
In conclusion, the 4A EDG being inoperable for an approximate 16.9 days, as a singular event, has minimal safety significance.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (61 PAGE (3) Turkey Point Unit 4 05000251
CORRECTIVE ACTION
A change to the maintenance procedure will be made to ensure that the run-out of the engine flywheel gear teeth is checked and that the magnetic pickups are set with adequate clearance from the top of the highest gear tooth.
ADDITIONAL INFORMATION
component function identifier (if appropriate)].
FAILED COMPONENTS IDENTIFIED:� None (not an intrinsic failure) PREVIOUS SIMILAR EVENTS:� None
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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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