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| Start date | Reporting criterion | Event description | |
|---|---|---|---|
| 05000382/LER-2017-002 | 17 July 2017 18 September 2017 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | On July 17, 2017, at 1606 CDT, Waterford 3 experienced an automatic reactor scram due to a loss of forced circulation, which was the result of a loss of off-site power to the safety and non-safety electrical busses. Prior to the scram, plant operators manually tripped the main turbine and generator due to overheating of the isophase bus duct due to the failure of a shunt assembly connection in the duct to Main Transformer 'B'. The automatic electrical bus transfer did not occur due to relay failures in the fast dead bus transfer system. Both 'A' and 'B' Emergency Diesel Generators started and loaded as designed to re-energize the 'A' and 'B' safety busses. The loss of off-site power caused a loss of both Main Feedwater pumps, resulting in an automatic actuation of the Emergency Feedwater system. The Root Cause of this event was the design change procedure used for modifications to the fast dead bus transfer circuitry did not include guidance to detect the susceptibility of the relays to DC coil inductive kick. The faulty relays in the fast bus transfer circuit were replaced prior to plant startup. An Unusual Event was declared at 1617 CDT due to loss of off-site power to safety buses for >15 minutes. All required safety-related equipment responded as expected during this event. |
| 05000382/LER-2017-001 | 8 March 2017 4 May 2017 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | On March 8, 2017, at 1627 CST, it was identified that Low Pressure Safety Injection (LPSI) train ‘B' was inoperable due to SI-135B, Reactor Coolant Loop 1 Shutdown Cooling Warmup Valve, being found open, which is not the required position. At the time of discovery, LPSI train ‘A' was inoperable for maintenance and the station was in compliance with Technical Specification (TS) 3.5.2 action ‘a' which requires that an inoperable LPSI train be restored within 7 days. The shift operating crew entered TS 3.5.2 action ‘c' due to both trains of the Emergency Core Cooling System being inoperable. Action ‘c' requires that with both LPSI trains inoperable, at least one train must be restored within one hour. SI-135B was subsequently closed and tested to verify operability. TS 3.5.2 action 'c' was exited at 1705. The station remained in compliance with TS 3.5.2 action ‘a'. It was determined that the SI-135B valve was opened inadvertently. It was planned to perform work on SI-135A, Reactor Coolant Loop 2 Shutdown Cooling Warmup Valve. The workers incorrectly began work on SI-135B and manually opened the valve. This was caused by personnel not performing proper component verification to validate that they were on the correct component, contrary to station procedures. Corrective actions are being performed to improve station work practices related to component verification. |
| 05000382/LER-2016-002 | 12 August 2016 30 November 2016 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | On August 12, 2016, at 1704 CDT, the shift operating crew noted that Essential Chiller B outlet temperature exceeded the allowed maximum Technical Specification (TS) Surveillance Requirement (SR) 4.7.12.1.b limit of 42 degrees Fahrenheit (deg F) and Essential Services Chilled Water Loop B (Loop B) was declared inoperable. Essential Services Chilled Water Loop A (Loop A) had previously been declared inoperable on August 11, 2016. The shift operating crew entered TS 3.0.3 due to both trains of Essential Services Chilled Water being inoperable. Essential Chiller AB was subsequently aligned to Loop B and TS 3.0.3 was exited on August 12, 2016, at 1802, when outlet temperature was verified to be less than or equal to 42 deg F. The Essential Chiller A elevated temperature was due to a failed capacity control module. The apparent cause was inadequate preventative maintenance (PM) strategy. Corrective action to replace the module and return Essential Chiller A to service is complete. In addition, the PM to replace the module will be performed more frequently. The Essential Chiller B elevated chilled water temperature was due to incorrect guide vane setup. The incorrect setup resulted in adjusting the thermostat too high. The high thermostat setting prevented the chiller from maintaining outlet temperature less than 42 deg F following a large demand increase. The apparent cause was that inadequate guidance exists for guide vane actuator linkage setup. Corrective actions to properly set up the linkage, set the thermostat, return Essential Chiller B to service, develop actuator installation instructions, and revise post- maintenance tests are complete. Additional actions include establishing new maintenance and troubleshooting procedures. |
| 05000382/LER-2016-001 | 3 July 2016 1 September 2016 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | On July 3, 2016, the Core Protection Calculator (CPC) Tilt Exceeded alarm was received. This required the Azimuthal Power Tilt values to be adjusted for the actual condition. During the adjustment of the Type 1 addressable constants, it was discovered that the CPC Tilt value for CPC C, PID 063, had been improperly inputted following a failure and restoration of the CPC C on July 1, 2016. The value entered was 1.0064 when the value should have been 1.0102. The inputted values were taken from the last known functional printout in the CPC C book but were not verified against the addressable constant change log and did not account for any subsequent changes made by operations. The value entered was lower than intended, and because of this low value, there were several times over the next 2 days that actual Azimuthal Power Tilt was greater than the value assumed by CPC C. This should have resulted in entry into TS 3.2.3.a. The plant operated for approximately 29 hours outside of the allowed TS actions per TS 3.3.1 and 3.2.3.a. This also resulted in CPC C Local Power Density (LPD) and Departure from Nucleate Boiling Ration (DNBR) calculating a non-conservative value. The Apparent Cause was that the CPC Functional Test procedure contains vague guidance on how to verify inputted addressable constants are correct. This resulted in this condition by implying to personnel that only using the CPC printed page of addressable constants to verify inputted addressable constants was correct and did not also prompt use of the CPC change log. Corrective actions include revising the procedure to provide specific guidance on how to verify inputted addressable constants are correct. |
| 05000382/LER-2014-004 | 22 October 2014 | 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 | During a walkdown of the Emergency Diesel Generator Feed Tank A and B vent lines on October 22, 2014, an NRC Component Design Basis Inspection inspector identified corrosion on the Emergency Diesel Generator Feed Tank A and B vent lines where the vent lines pass through the roof. A visual inspection was performed and revealed that the corrosion had created through wall holes that could allow water into both the train A and B Emergency Diesel Generator Feed Tanks. Follow up analysis has determined that some rainfall amount less than the postulated Probable Maximum Precipitation event could have resulted in water intrusion into the Emergency Diesel Generator A and B Feed Tanks that exceeds the 0.1 percent water content allowed by the vendor technical manual. This could have potentially affected the operability of both the A and B Train Emergency Diesel Generator Feed Tanks and subsequently both trains of the Emergency Diesel Generators. It is unknown how long this corrosion has existed. Compensatory measures were put in place to prevent water ingress should a large rainfall event occur. This condition is reportable under the following criteria: 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(v)(D), and 10 CFR 50.73(a)(vii). |
| 05000382/LER-2014-003 | 18 August 2014 | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | During a planned system outage of train B Component Cooling Water (CCW), which rendered train B Ultimate Heat Sink (UHS) inoperable, there was an unexpected trip of train A Auxiliary Component Cooling Water (ACCW) Wet Cooling Tower (WCT) fan, which rendered the train A UHS also inoperable. This resulted in both trains of the UHS system inoperable for approximately 83 minutes. The ACCW WCT Fan 6A tripped at approximately 0853, rendering the redundant train A UHS inoperable, causing entry into 1 hour TS LCO 3.7.4 ACTION (b), which requires restoring at least one UHS train to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. Then at approximately 0948 the WCT Fan 6A electric motor thermal overload relays were reset, the fan restarted, and operated properly. At 1016, CCW train B had been restored from the planned maintenance and was declared operable, exiting TS LCO 3.7.4 ACTION (b). The direct cause to the WCT fan 6A trip was due to localized heating at the motor starter T-lead connection on the thermal overload. Technicians found the lug was tight (galled) but the motor T-lead showed looseness when moved. The lug was replaced. The cause was determined to be the procedure contained insufficient detail on how to verify successful implementation for tightening a mechanical lug, and corrective action is planned to add the appropriate detail. |
| 05000382/LER-2014-002 | 1 March 2014 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | During an operator tour of the train A Emergency Diesel Generator (EDG) room on March 1, 2014 at approximately 12:39, it was discovered that the filter housing cover on the EDG A Starting Air Filter had unfastened from its base and was lying on the floor. The last successful start of the EDG A that demonstrated its capability to meet its safety function was on February 27, 2014 at 11:33 and it is postulated that the cover became unfastened during that run. Therefore, EDG was potentially inoperable for 2 days as a result of this condition. Since the condition was unknown at the time, the Technical Specifications required test of EDG B was not performed within 8 hours and the requirement to demonstrate the operability of the remaining A.C. circuits at least once per 8 hours or to be in Hot Shutdown within the next 6 hours was not performed. The cause was insufficient tightening of the filter housing cover during maintenance on April 8, 2013. The insufficient tightening had not been preventing the EDG A's ability to start within the required times prior to its becoming unfastened. EDG A was returned to OPERABLE condition by refastening the filter housing cover to its base. The other filter housings on EDG A and EDG B were verified to be properly fastened. An evaluation is being conducted to determine if a torque requirement is necessary for the filter housing covers. |
| 05000382/LER-2014-001 | 8 May 2013 | 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 | On May 8, 2013 at 19:53, a safety-related circuit breaker failed when Operations personnel attempted to start Shutdown Cooling Heat Exchanger Room B air handling unit. In accordance with Operations administrative procedure for Technical Specification and Technical Requirements Compliance, Operations personnel entered the applicable Technical Specification Limiting Condition for Operation for Containment Spray (CS) train B. A replacement breaker was installed and a successful start of the air handling unit was performed. CS Train B was declared operable at 17:30 on May 9, 2013. An evaluation of the failure determined that the air handling unit had been effectively rendered inoperable since installation of the circuit breaker on April 18, 2013. Containment Spray train B was inoperable from 03:09 on April 17, 2013, when the train was declared inoperable to perform preventative maintenance until 17:30 on May 9, 2013, a total of 22.6 days. This time period exceeds the Technical Specification allowed outage time of 7 days. Technical Specification 3.6.2.1 was not complied with, which requires that with one CS system (train) inoperable, restore the system (train) to operable status within 7 days or be in at least Hot Standby within the next six hours. During this time frame CS 'A' was inoperable for 6.65 hours. Since both trains of Containment Spray were inoperable this is considered a Safety System functional failure due to the system being unable to mitigate the consequences of an accident. |
| 05000382/LER-2013-007 | 18 October 2013 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | On October 18, 2013, personnel at Waterford 3 identified that, contrary to Technical Specification (TS) 3.9.7, during dry fuel loading activities in the Fuel Handling Building, with the transfer cask loaded with fuel, a load of over 2000 pounds was placed over the transfer cask with a crane that is not single failure proof. This condition occurred with the irradiated fuel covered with a stainless steel lid, 9.5 inches thick, welded in place covering the fuel. TS 3.9.7 Limiting Conditions of Operation (LCO) requires loads in excess of 2000 pounds to be prohibited from travel over irradiated fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using a single-failure-proof handling system. This condition is reportable under 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by TS. The procedure governing this activity was changed to require using the single-failure proof handling system of the FHB crane main hook instead of allowing the use of FHB auxiliary hook that is not considered single-failure proof. |
| 05000382/LER-2013-006 | 4 September 2013 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | On September 4, 2013, personnel at Waterford 3 determined that the manual operator for a safety related air operated temperature control valve (ACC-126A) could not perform a full manual closure of the valve. Automatic operation was not affected. Manual closure is credited in safety analysis to preserve wet cooling tower inventory as part of the ultimate heat sink. The causal review determined that the cause, misalignment of the valve disc to the shaft, had occurred during planned maintenance in July of 2012. During this term of past inoperability, required systems and components in the redundant safety train were also inoperable for short periods of time for surveillance and planned maintenance reasons. The valve was repaired, retested successfully, and declared operable on September 5, 2013. There was no impact to nuclear safety due to this condition. This condition is reportable under 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(ii)(B). |
| 05000382/LER-2013-005 | 22 May 2013 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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)(C), 50.54(x) TS Deviation | Waterford 3 declared Emergency Diesel Generator B (EDG-B) inoperable on May 22, 2013 due to inability to maintain room temperature within design limits. Subsequent trouble shooting revealed that the variable pitch room exhaust fan had failed due to separation of the fan hub from the hub sleeve. Examination of recent operating data showed that the first evidence of fan failure had been during a surveillance test the previous month. An apparent cause evaluation determined the probable cause of the failure to be the result of repairs made during a previous (1999) fan motor replacement. These repairs caused additional stresses on the fan hub components which eventually resulted in fan hub separation from the hub sleeve. The EDG-B room exhaust fan was repaired and EDG-B operability was restored on May 26, 2013. Safety significance for the event is characterized as low to moderate. This condition is reportable under the following criteria: 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(ii)(B), and 10 CFR 50.73(a)(2)(v)(D). |
| 05000382/LER-2013-004 | 29 April 2013 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(C), 50.54(x) TS Deviation | A review of plant reactor coolant temperature trends from a planned outage shows that on April 29, 2013, reactor coolant temperature was raised from approximately 135 degrees F to approximately 180 degrees F as a planned evolution. On April 30, 2013, RCS temperature was raised from approximately 180 degrees F to approximately 345 degrees F as a planned evolution. Additionally, on April 30, 2013, plant operations involving water additions to the Volume Control Tank were conducted. A review of the Station Log shows that only one ENI log channel was operable during these evolutions. These evolutions were prohibited by the plant's Technical Specifications under these conditions. TS 3.3.1 Action 4 requires, with only one log power channel operable in Mode 5 or Mode 4, suspending "all operations involving positive reactivity changes," and is clarified by an applicable note which states, "Limited plant cooldown or boron dilution is allowed provided the change is accounted for in the calculated shutdown margin. Planned corrective action is to revise the TS bases for TS 3.3.1 to emphasize the limitations for usage of the note applicable to Action 4. The planned heatup and dilution evolutions had no discernable effect on nuclear safety. |
| 05000382/LER-2013-003 | 16 November 2012 | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(b) 10 CFR 50.73(a)(2)(ix) 10 CFR 50.73(a)(2)(i)(C), 50.54(x) TS Deviation | On November 16, 2012 at 03:29 CST, it was identified that periodic testing had not been established for the local manual handwheel function on twenty-four safety related Air Operated Valves (AOV) that are required by design to operate after their associated air supply accumulator is exhausted. Additionally, license basis documents had not been updated to reflect an accumulator mission time. An inadequate evaluation had been performed of the substitution of manual operator action for automatic action by the AOV. Testing revealed that three AOVs would not operate by manual handwheel. There was one inoperable AOV in each of the Emergency Feedwater (EFW) flow paths to the Steam Generators, which made both EFW trains inoperable. There was an inoperable AOV manual override in the Auxiliary Component Cooling Water System, which made and the Train B Ultimate Heat Sink (UHS) inoperable. Since testing of the three AOV's handwheels had not been performed, it is indeterminate how long the condition existed; however, the condition had recently existed where both trains of EFW and UHS were required to be OPERABLE. At the time of discovery, Waterford 3 was defueled in a no mode condition and was in compliance with Technical Specification requirements. Actions were completed to restore the three valves' manual handwheels to an operable status. |
| 05000382/LER-2013-002 | 21 January 2013 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | On January 21, 2013 at 15:51 CST, Waterford Steam Electric Station, Unit 3 (Waterford 3) experienced a reactor trip from approximately 91% power due to lowering Steam Generator (SG) #1 level. Emergency Feedwater (EFW) Actuation Signals (EFAS-1 and EFAS-2) were received due to low SG levels, which is an anticipated response to the reactor trip with the plant at or near full power. EFW flow to SG 1 was identified to be oscillating from 0 to 800 GPM. While resetting EFAS, EFW flow to SG 1 was approximately 120 GPM without a demand signal. Valve EFW-223A (Emergency Feedwater to SG1 Backup Flow Control Valve) was declared inoperable. Subsequent evaluation determined that EFW-223A past operability was affected by the apparent cause. Plant response to the reactor trip was unaffected by the condition and all design safety function capabilities remained available. This condition did not compromise the health or safety of the general public. The past inoperability of valve EFW-223A is reportable as a Licensee Event Report pursuant to 10CFR50.73(a)(2)(i)(B), Operation or Condition Prohibited by Technical Specifications. NRC FORM 3E6 (10-2010)obo3 |
| 05000382/LER-2013-001 | 21 January 2013 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | On January 21, 2013 at 15°51 CST, Waterford 3 experienced an automatic reactor trip from approximately 91% power due to lowering Steam Generator (SG) #1 level following the unexpected closure of Main Feedwater Regulating Valve #1 due to an instrument air line failure. Emergency Feedwater (EFW) Actuation Signals (EFAS-1 and EFAS-2) were received due to low SG levels, which is an anticipated response to the reactor trip with the plant at or near full power. SG #1 received EFW system flow for a short period of time. The plant stabilized in Mode 3 with levels in both SG's restored to normal operating band with the Main Feedwater (MFW) system. Adequate water level was maintained in the SG's during the transient to ensure decay heat removal from the Reactor Coolant System (RCS). This condition did not compromise the health and safety of the general public. This condition is reportable pursuant to 10CFR50.73(a)(2)(iv)(A) due to the automatic actuation of the Reactor Protection System (RPS) and due to the automatic actuation of the EFW system. |
| 05000382/LER-2012-005 | 1 June 2012 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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 | On 6/1/2012 at 04:26 CDT, Operations declared Auxiliary Component Cooling Water (ACCW) Train A inoperable due to excessive seat leakage through air operated temperature control valve ACC-126A. The Component Cooling Water Technical Specification 72 hour shutdown action was entered. The valve was unable to meet its low leakage safety function when closed by flow demand. Following maintenance, ACCW Train A was declared operable on 6/2/2012 at 14:30 CDT. Similar conditions have occurred since October of 2011. Engineering evaluation has determined the cause to be wearing of the valve stem bushings for ACC-126A. Since the condition is only recognized when placing the system in standby, the recurring condition created several occasions where past operability of ACCW Train A was not assured. Train B has not exhibited this problem. Analysis of past operability has determined that the impact on safety significance of ACC-126A failing to fully close is minimal. This condition is being reported pursuant to the requirements of 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(ii)(B), 10 CFR 50.73(a)(2)(v)(D), and 10 CFR 21. |
| 05000382/LER-2012-001 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | During the recent NRC Triennial Heat Sink Performance Inspection, it was identified that Component Cooling Water flows to the B and D Containment Fan Coolers during normal operating conditions were below the Technical Specification (TS) Surveillance Requirement (SR) for minimum flow from approximately July 8, 2009 through July 19, 2009. The duration of this condition exceeded the TS 3.6.2.2 Limiting Condition for Operation Allowable Outage Time of 72 hours. TS SR 4.6.2.2.a requires that each train of containment cooling shall be demonstrated operable at least once per 31 days by verifying a cooling water flow rate of greater than or equal to 625 GPM to each cooler. The original condition that caused the flows to be less than 625 GPM during normal operating conditions has since been corrected and subsequently meets or exceeds the flow-rate listed in the surveillance requirement. No plant transient or safety system actuations occurred. Stable plant operation continued at 100 percent power. | |
| 05000382/LER-2011-005 | 20 October 2011 | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | At 08:00 CDT on 10/20/2011, Essential Chilled Water Loop B was declared inoperable due to equipment failure while turbine driven Emergency Feedwater (EFW) Pump AB was out of service for planned maintenance. Operability of Essential Chilled Water Loop B was restored at 08:50. During this time period, the application of cascading technical specifications rendered motor driven EFW Pump B inoperable. The remaining operable EFW Pump A is a design rated 50 percent pump; therefore, this event could have prevented fulfillment of the residual heat removal safety function. Offsite power and Train A safety related equipment and systems were verified operable. Essential Chilled Water Loop B was declared inoperable because Essential Chiller B failed to automatically restart from a load recycle. The Operations crew took immediate action to align Essential Chiller AB to restore operability to Essential Chilled Water Loop B. Restoration of operability to Essential Chilled Water Loop B concurrently restored operability to EFW Pump B. No plant transient or safety system actuations occurred. Stable plant operation continued at 100 percent power. |
| 05000382/LER-2011-004 | 27 July 2011 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | On July 27, 2011, the Waterford 3 Steam Electric Station (W3) received a Westinghouse Nuclear Safety Advisory Letter which informed W3 that a power supply harness supplied for the Qualified Safety Parameter Display System (QSPDS) may be undersized and not be able to handle the current to which it is exposed under all environmental conditions. W3 determined that the suspect harness had been installed in both QSPDS channels. No actual failure occurred. Technical Specification (TS) 3.3.3.6 requires the accident monitoring instrumentation channels (including QSPDS) to be OPERABLE in MODES 1, 2, and 3. The identified condition rendered both channels of QSPDS inoperable. TS 3.3.3.6 Action 30 specifies an allowed outage time (AOT) of seven days. The same vendor letter recommended installation of a jumper to restore operability. This was performed by site maintenance personnel to both QSPDS channels on July 29, 2011. This condition is reportable under 10CFR50.73(a)(2)(i)(B), operation or condition prohibited by Technical Specifications, because the suspect power supply harness was installed in both channels for greater than the TS AOT. |
| 05000382/LER-2011-003 | 30 April 2011 | 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 | On April 30, 2011, the Emergency Diesel Generator (EDG) A output breaker failed to automatically close during Technical Specification (TS) surveillance testing. This was caused by a human performance error on November 2, 2010 when an electrical lead was reconnected to the wrong terminal on a timing relay following maintenance. With this incorrect configuration, EDG A was not able to automatically energize the Train A electrical safety bus. Operator action would have been required to energize the bus by locally closing the EDG A output breaker. Waterford 3 was in mode 5 at the time of discovery, and was in compliance with TS requirements with EDG B operable. Due to EDG A being inoperable from November 2, 2010 to after the time of discovery, the TS 3.8.1.1 Limiting Conditions for Operation (LCO) requirements were not met. Additionally, Train B equipment had been inoperable during this period which presented a condition that could have prevented the fulfillment of a safety function. The wiring error was corrected on May 1, 2011 and EDG A was restored to service on May 2, 2011. Planned corrective actions include procedure changes that will require the use of plant design documents to verify as left conditions and strengthening post maintenance testing. |
| 05000382/LER-2011-002 | 7 April 2011 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | On 4/7/2011, Main Feedwater Isolation Valve (MFIV) "A" failed the in-service test (1ST) stroke time. Due to the failed stroke time, the plant operated in a condition prohibited by Technical Specifications. Technical Specification 3.7.1.6 requires MFIV "A" to be operable in Modes 1 through 4, with a surveillance requirement 4.7.1.6 that requires the closure time of less than or equal to 6 seconds. MFIV "A" uses a hydraulic actuator, which contains two separate hydraulic accumulators. An actuation signal results in both accumulators closing the valve. The Root Cause Evaluation determined that the hydraulic actuator four way valves are marginally designed to provide the motive force needed to overcome infrequent valve operation coupled with the condition that Fyrquel hydraulic fluid is susceptible to breakdown. Corrective actions included replacing the four way valves and replacing the hydraulic fluid for MFIV "A". |
| 05000382/LER-2011-001 | 16 February 2011 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | This LER reports the operation of Waterford 3 in a condition prohibited by Technical Specification 3/4.11.2.5, Radioactive Effluents, Explosive Gas Mixture. Specifically, Technical Specification 3/4.11.2.5 requires that the concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2 percent by volume at all times whenever the hydrogen concentration exceeds 4 percent by volume. At 4:10 p.m. on February 14, 2011, it was identified that Gas Decay Tank C oxygen concentration had exceeded the allowed concentration for the existing hydrogen concentration specified in Technical Specification 3.11.2.5. Technical Specification 3/4.11.2.5 Action 'a' states in the event that oxygen concentration is greater than 2 percent by volume, but less than 4 percent by volume, oxygen concentration must be reduced to 2 percent or less within 48 hours. Contrary to this requirement, a review of the condition determined that the 48 hour action limit specified in Technical Specification 3/4.11.2.5 Action 'a' was exceeded and is reportable. The oxygen concentration was returned to within limits at 10:47 p.m. on February 18, 2011, and compliance with Technical Specification 3/4.11.2.5 was restored. There have been no previous similar licensee events reported in the last three years. |
| 05000382/LER-2010-003 | 8 February 2010 | 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 | On February 8, 2010, at approximately 14:47, with the plant operating at 100% power (Mode 1), Waterford 3 commenced a post-surveillance cooldown on Emergency Diesel Generator A (EDG A) (EK). The local Nuclear Auxiliary Operator noted the one inch main fuel oil supply line support clamp was loose and the supply line tubing had circumferential wear indication under the clamp. An engineering review of the wear indication found the wall thickness at 0.0132" remaining from a nominal wall'thickness of 0.060". An Engineering calculation determined a tubing wear rate that showed a design basis 30 day run by the EDG A would not have been met since approximately August 31, 2005 without mitigating actions.)Since August 2005, there were several occurrences when EDG B was also out of service. This condition represents operation not in compliance with Technical Specifications and a safety system functional failure. Additionally, Technical Specification (TS) requirements were not met when the steam driven EFW pump was periodically unavailable while EDG A was in this condition. The tubing was replaced on February 10, 2010. The event did not compromise the health and safety of the general public. |
| 05000382/LER-1999-014, Forwards LER 99-014-00,providing Details of Reactor Shutdown Due to Loss of RCP Controlled bleed-off Flow.Attached Commitment Identification/Voluntary Enhancement Form Identifies All Commitments Contained in Submittal | 12 October 1999 | ||
| 05000382/LER-1999-013, Forwards LER 99-013-00,providing Details of Exceeding TS Limits for RCS Cooldown Rates.All Commitments Contained in Submittal Are Identified on Encl Commitment Identification/ Voluntary Enhancement Form | 23 September 1999 | ||
| 05000382/LER-1999-011, Forwards LER 99-011-00,providing Details of Reactor Shutdown Due to Loss of Controlled bleed-off Flow.All Commitments Contained in Submittal Identified on Attached Commitment Identification/Voluntary Enhancement Form | 31 August 1999 | ||
| 05000382/LER-1999-010, Forwards LER 99-010-00,providing Details of Inadequate Pumping Capacity in Dry Cooling Tower Area.All Commitments Contained in Submittal Are Identified on Attached Commitment Identification Voluntary Enhancement Form | 26 August 1999 | ||
| 05000382/LER-1999-009, Forwards LER 99-009-00 Re Discovery of Condition of Noncompliance with App R Involving Inadequate Separation of Essential Cables Routed in Fire Area RAB-30 in Rab. Compensatory Measures Were Established Immediately | 26 August 1999 | ||
| 05000382/LER-1999-008, Forwards LER 99-008-00,re Failure to Perform Testing of ESF Filtration Units Per TS Srs.Commitments Made by Util Also Encl | 29 July 1999 | ||
| 05000382/LER-1999-007, Forwards LER 99-007-00,providing Details of Operation Outside Tornado Missile Protection Licensing Basis for turbine-driven Emergency Feedwater Pump Exhaust Stack & Steam Supply Piping.All Commitments Identified on Attached | 23 July 1999 | ||
| 05000382/LER-1999-005, Forwards LER 99-005-00,providing Details of Discovery of Untested Electrical Contacts in safety-related Logic Circuits | 24 June 1999 | ||
| 05000382/LER-1999-004, Forwards LER 99-004-00 Re Discovery That Response Time Testing Had Not Been Performed for ESFAS Containment Cooling Function,As Required by TS SR 4.3.2.3 | 14 May 1999 | ||
| 05000382/LER-1999-002, Forwards LER 99-002-00 Re Leaks Discovered on Instrument Nozzles in Rcs.Event Being Reported Per 10CFR50.73(a)(2)(ii) & 10CFR50.73(a)(2)(i)(B) | 25 March 1999 | ||
| 05000382/LER-1998-021, Forwards LER 98-021-00,providing Details of Condition in Which Plant Was Operated with B Concentration in One of SITs Potentially Below TS Limits.Reasonable Assurance Exists That Condition Existed Over Period of Time Greater th | 11 January 1999 | ||
| 05000382/LER-1998-019, Forwards LER 98-019-01,IAW 10CFR50.73(a)(2)(ii). LER 98-019-00,discussed.Refs to Feedwater Valves Have Been Deleted in Revised LER | 8 December 1998 | ||
| 05000382/LER-1998-018, Forwards LER 98-018-00 for Waterford Steam Electric Station, Unit 3.Rept Provides Details of Situation in Which TS 3.0.3 Entered,Due to Both Trains of Shield Bldg Ventilation Being Inoperable Simultaneously | 1 October 1998 | ||
| 05000382/LER-1998-017, Forwards LER 98-017-00,re Error in Energy Redistribution Factors Used in LOCA Analysis.Condition Was Also Reported to NRC Via Telcon 980803 | 2 September 1998 | ||
| 05000382/LER-1998-015, Forwards LER 98-015-00,providing Details of Incorrect Installation of T-ring Rubber Seals on CR Normal Outside Air Intake Upstream & Downstream Isolation Valves HVC-101 & HVC-102 | 24 August 1998 | ||
| 05000382/LER-1998-014, Forwards LER 98-014-00 for Waterford Steam Electric Station, Unit 3.Rept Provides Details of Manual Reactor Trip,Followed by Actuation of EFW Actuation Sys That Initiated EFW Flow to Both SGs | 17 August 1998 | ||
| 05000382/LER-1998-012, Forwards LER 98-012-00,re Details of Condition Which Constituted Violation of TS 3.3.1 Because TS Channel Check Was Not Performed,Per TS Requirement 4.3.1.1 | 9 July 1998 | ||
| 05000382/LER-1998-011, Forwards LER 98-011-00,providing Details of Condition Wherein Plant Entered TS 3.0.3 for Having Both Trains of Component Cooling Water Declared Inoperable Concurrently | 2 July 1998 | ||
| 05000382/LER-1998-010, Forwards LER 98-010-00,re Potential Overpressurization of Nitrogen Sys Possibly Rendering Multiple Safety Sys Inoperable | 15 June 1998 | ||
| 05000382/LER-1998-009, Forwards LER 98-009-00,providing Details of Inadvertent ESF Actuation Which Started CCS Components.Rept Submitted Per, 10CFR50.73(a)(2)(iv) | 15 May 1998 | ||
| 05000382/LER-1998-008, Forwards LER 98-008-00,re Degraded Condition in Both Trains a & B of Ecw Sys.No New Commitments Made within Ltr | 14 May 1998 | ||
| 05000382/LER-1998-006, Forwards LER 98-006-00 Re Discovery of Feedwater Isolation Valves Being Inoperable in Excess of TS Allowed Outage Time. Condition Being Reported Pursuant to 10CFR50.73(a)(2)(i)(B) | 13 April 1998 | ||
| 05000382/LER-1998-005, Forwards LER 98-005-00,re Details of Missed TS 3.8.4.2 Surveillance Test for Thermal Overloads of Eight motor-operated Valves.Condition Being Reported Pursuant to 10CFR50.73(a)(2)(i)(B) | 13 April 1998 | ||
| 05000382/LER-1998-004, Forwards LER 98-004-00,providing Details of TS 4.0.3 Condition That Existed W/Ts 4.3.1.1,Table 4.3-1,refueling Interval Channel Functional Test for Core Protection Calculator Sys | 3 April 1998 | ||
| 05000382/LER-1998-003, Forwards LER 98-003-00 Providing Details of Both Hydrogen Analyzers Being Inoperable Due to Undersized Thermal Overloads.Condition Being Reported Per 10CFR50.73(a)(2)(i) (B) & 10CFR50.36(c)(2) Prohibited by TSs | 1 April 1998 | ||
| 05000382/LER-1998-002, Forwards LER 98-002-00 Providing Details of Condition Involving Brief Loss of Command Function in CR in Violation of Administrative TS 6.2.2.Condition Being Reported Per 10CFR50.73(a)(2)(i)(B) | 18 March 1998 | ||
| 05000382/LER-1998-001, Forwards LER 98-001-00,providing Details of TS 3.0.3 Condition Experienced During Plant Malfunction When LCO Section 3.1.2.4 Was Not Met | 16 March 1998 |