|Report date||Site||Event description|
|05000219/LER-2017-004||16 February 2018||Oyster Creek|
On August 31, 2017, during a review of industry Operating Experience (FERMI 2, LER 2017-001) for the use of a Reactor Protection System (RPS) test box during main turbine surveillance testing, it was determined that Oyster Creek station procedures failed to implement the required action specified by Technical Specifications (TS) section 3.1,1. note (nn) during testing. The surveillance tests associated with the Turbine Trip and Generator Load Rejection functions were revised in 2013 to use an RPS test box in order to minimize operational risks associated with the receipt of half scram signals during testing. The installation of the RPS test box caused two of the four required instrument channels for the Turbine Trip Scram function to be bypassed during testing.
In accordance with station Technical Specifications, the required action to verify sufficient channels remained operable was not documented as complete within the action time specified in TS Table 3.1.1, note (nn). This issue was identified under normal operating conditions, and is reportable under 10 CFR 50.73(a)(2)(i)(B).
|05000219/LER-2017-005||3 January 2018||Oyster Creek|
On 10/09/17, during the bi-weekly Emergency Diesel Generator (EDG) #2 Load Test, a Generator Lockout signal (86G) was received which tripped the EDG output breaker. This failure resulted in EDG #2 being declared inoperable, and entry into an unplanned 7-day Limiting Condition for Operation (LCO) at 0312.
Troubleshooting identified a broken electrical ring lug connector on a current transformer that provides an input to the protective relay logic. The investigation determined the connector failure was due to fatigue cracking that was initiated by stresses caused by bending and twisting of the electrical lug beyond limits specified in industry guidelines. The ring lug was most likely distressed during initial installation in the 1990's. The condition that led to the EDG #2 trip existed for greater than the allowed outage time in the plant's Technical Specification (TS) and is reportable under 10 CFR 50.73(a)(2)(i)(B).
|05000219/LER-2017-003||31 August 2017||Oyster Creek|
On July 3, 2017 at approximately 10:30 hours with the reactor subcritical after a manual scram, an automatic reactor scram occurred due to low Reactor Pressure Vessel (RPV) water level. Prior to this event, Main Control Room (MCR) Operators established reactor water letdown, reset the manual scram and were placing RPV water level control in automatic using a Low Flow Feedwater Regulator Valve (LFRV). Reactor pressure control was automatically being maintained with turbine bypass valves. A low reactor water level occurred when a bypass valve opened as expected to lower reactor pressure.
Operations personnel had reset the scram signal without fully evaluating other plant conditions. Upon resetting the scram with reactor water level not yet stabilized, an automatic scram was received on low reactor water level. The automatic scram initiation signal occurred with the plant in a shutdown mode.
|05000219/LER-2017-002||31 August 2017||Oyster Creek|
On July 3, 2017, at approximately 10:15 AM following a grid disturbance, a manual scram was inserted due to degrading main condenser vacuum because of an improper configuration of the Augmented Off-gas (AOG) System.
The loss of main condenser vacuum resulted when Operations personnel failed to execute procedural requirement to align the AOG system into a shutdown lineup. The loss of vacuum was caused by degraded Steam Jet Air Ejectors (SJAE) performance due to a blocked discharge path.
The AOG system tripped 11 hours earlier following a grid disturbance. During the trip, operations personnel failed to re-align the AOG treatment system to a shutdown lineup resulting in the AOG Flame Arrestor siphoning into the inlet piping which filled the lower section of the off-gas hold up line with water.
|05000219/LER-2016-001||24 January 2017||Oyster Creek|
On January 4, 2016, during the biweekly load test surveillance, the Emergency Diesel Generator (EDG) #1 tripped after 26 minutes of operation and an unexpected Main Control Room alarm was received.
Locally, at the EDG enclosure, it was discovered that the rubber pipe coupling that connects the coolant surge tank to the right bank cooling water pump inlet tee had ruptured and that the EDG #1 control circuitry processed the trip on low coolant pressure.
EDG #1 was declared inoperable and in accordance with Technical Specification (TS) 3.7.C.2 Limiting Condition for Operation (LCO), the unit entered a seven day action period. The EDG #1 was returned to service on January 5, 2016, at 0130 hours, following successful hose replacement and completion of the load test surveillance.
The Root Cause investigation determined that an activity to replace the rubber pipe coupling on a 12-year frequency was not established during the initial implementation of Exelon's Electro-Motive Division Diesel Generator Performance Centered Maintenance (PCM) template.
This event resulted in an Operation or Condition that was Prohibited by the Plant's Technical Specifications and is therefore being reported under 10CFR50.73(a)(2)(i)(B).
|05000219/LER-2016-005||17 November 2016||Oyster Creek|
On September 19, 2016, after achieving Cold Shutdown for the 1R26 Refuel Outage, as found testing was performed on all five (5) Electromatic Relief Valves (EMRVs). The "E" EMRV did not open from the Main Control Room (MCR), and no change in indication was observed. Per the work activity, technicians were dispatched to the Drywell to verify that the valve did not move upon receiving an open signal from the MCR.
A cutout switch in the valve actuator was stuck in the open position, thereby preventing the solenoid from actuating to open the valve. The cutout switch did not operate as required due to, hinge pin washers not installed in the cutout switch assembly. Without the washers installed to the hinge pins interfered with the solenoid frame holes creating mechanical binding. Based on this information it is suspected that the "E" EMRV would have been inoperable for longer than Technical Specification Allowed Out of Service Time (AOT) of 24 hours. Testing and inspections were performed on all EMRVs prior to installation in the plant.
Therefore, this issue is reportable under 10 CFR 50.73(a)(2)(i)(B) as an Operation or Condition which was Prohibited by the plant's Technical Specifications.
|05000219/LER-2016-003||29 June 2016||Oyster Creek|
On April 30, 2016, at 1804 hours, during the plant startup following the 1M38 maintenance outage, a reactor SCRAM was manually inserted by the Control Room Operators during post maintenance testing following work on the 'D' Reactor Recirculation Pump (RRP) mechanical seal. The SCRAM was initiated since leakage was discovered by a rising trend in drywell unidentified leakage during plant startup. The seal had been replaced during the maintenance outage. The SCRAM was selected as the preferred method of shutting down the reactor due to low decay heat conditions following the outage.
ENS 51895 was submitted on April 30, 2016, as required by 10 CFR 50.72 (b)(2)(iv)(B). This issue is reportable under 10 CFR 50.73(a)(2)(iv)(A), for any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph 10 CFR 50.73(a)(2)(iv)(B).
|05000219/LER-2016-002||12 May 2016||Oyster Creek|
On 03/16/2016, it was identified that isolating or reducing cooling water to the Hydraulic Control Units (HCUs) for three control rods should have been considered a modification since it had the potential to impact the scram times of the control rods. Even though scram time penalties were applied for the three control rods where the cooling water flow was either isolated or reduced, failing to identify the isolation of cooling water to the control rods as a modification as described by the Technical Specifications resulted in the Plant not taking the action to scram time test the affected rods.
By not completing scram time testing for the control rods, whose cooling water was isolated or reduced, the station was in violation of the requirements of Technical Specifications Section 3.2, since the issue was not identified previously and the affected control rods were not declared inoperable and isolated.
|05000219/LER-2015-001||12 August 2015||Oyster Creek|
On March 22, 2015 at 1414, an automatic SCRAM from full power operation occurred at Oyster Creek due to a valid RPS actuation on APRMI-11.1-liflux. The APRM Iii-Hl flux was caused by a rise in reactor pressure due to the failure of the Electric Pressure Regulator (EPR). The backup Mechanical Pressure Regulator (MPR) did not limit reactor pressure.
The scram occurred due to inadvertent contact with degraded wiring during troubleshooting to isolate a DC ground identified within the turbine control indication circuitry. The contact caused a loss of signal from the EPR to its controlling Programmable Logic Controller (PLC) resulting in a loss of the EPR. This loss of signal, concurrent with the MPR being out of position, resulted in the subsequent rise of both reactor pressure and APRM flux that caused the SCRAM.
ENS 50916 was submitted on March 22, 2015 as required by 10 CFR 50.72 (b)(2)(iv)(B). This issue is reportable under 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B).
NRC FORM 388 (01-2014) Nq LICENSEE EVENT REPORT (LER)
|05000219/LER-2014-005, Secondary Containment Declared Inoperable||11 August 2015||Oyster Creek|
On September 19, 2014, -during refueling outage 01R26 with the unit in cold shutdown, a technician discovered that a previously authorized and installed Temporary Configuration Change (TCC) had been removed from a penetration in the Outboard Main
ENS 50476 was submitted on September 20, 2014. This issue is reportable under 10 CFR 60.73(a)(2)(v)(C) as an event that could have prevented the fulfillment of the safety function of a system needed to control the release of radioactive material, and 10 CFR 50.73(a)(2)(I)(8) as a condition which was prohibited by the OM'S Technical Specifications.
|05000219/LER-2014-004||17 November 2014||Oyster Creek|
On September 18, 2014, during refueling outage 01R25 with the unit in cold shutdown, plant personnel performed local leak rate testing (LLRT) of the Main Steam Isolation Valves (MSIVs). During the testing, it was identified that the "B" main steam line inboard isolation valve NSO3B (V-1-8) exceeded the leakage rate limits allowed by the plant's Technical Specifications.
The MSIV (V-1-8) that exhibited excessive leakage was repaired and the as-left leakage rate was verified to be within Technical Specifications leakage limits.
This issue is reportable under 10 CFR 50.73(a)(2)(i)(B) as a condition which was prohibited by the plant's Technical Specifications.
|05000219/LER-2014-001||8 September 2014||Oyster Creek|
On July 11, 2014 at approximately 0312 EDT, during planned reactor power ascension with reactor power at approximately 56% of rated thermal power, main condenser vacuum began to degrade. In accordance with the abnormal operating procedure for degrading vacuum, Operators inserted a manual scram of the reactor at 0314 EDT.
Following the reactor scram, operations and maintenance personnel identified an approximate 2"x6" hole and an approximate 2”x3” hole on the last convolute of the downstream side of Y-1-26 ('B' Condenser Steam Inlet Expansion Joint). It was confirmed to be an active leak and subsequently the source of condenser vacuum degradation. As a corrective action to this event, the expansion joint was replaced.
All control rods fully inserted and plant response was as expected. This event is being reported pursuant to 10CFR50.73(a)(2)(iv)(A) due to an actuation of the Reactor Protection System (RPS).
|05000219/LER-2013-005||11 April 2014||Oyster Creek|
On December 17, 2013 at 1958 EST, while shut down, the plant experienced a reactor SCRAM when taking the Mode Switch from REFUEL to SHUTDOWN. The jumpers required to prevent a full SCRAM for this Mode Switch change were not installed as required by procedure. The actuation was a result of the reactor mode switch being placed from the refuel position to the shutdown position without the scram bypass jumpers installed.
The root cause determined that an unvalidated assumption by the supervisors resulted in a reactor scram while shut down.
Corrective actions include evaluation on requirements, through an Out-of-box Evaluation, to ensure operators understand and can demonstrate the proper use of the human performance fundamental tools that broke down in this instance. In addition, procedures will be enhanced to clarify verification requirements.
There were no safety consequences impacting the plant or public safety as a result of this event. The reactor was subcritical with all rods inserted at the time of the actuation. All systems functioned as designed. This event is being reported pursuant to 10CFR50.73(a)(2)(iv)(B) due to a valid actuation of the Reactor Protection System (RPS).
|05000219/LER-2013-004||12 February 2014||Oyster Creek|
On 12/14/13 at approximately 0336 EST, during quarterly turbine valve testing with reactor power at 95% of rated thermal power, the plant experienced reactor pressure control abnormalities. Turbine Control Valves 2 and 3 failed closed due to the Servo Motor Feedback Support Bracket bolts backing out and falling out thereby requiring a scram. Operators initiated a manual reactor scram due to reactor pressure rising to 1042 psig which approached the scram set point.
These conditions were corrected during 1F33. These were determined during complex troubleshooting as the failures that drove the event. The root cause determined that the manufacturer failed to assemble the Control Valve Hydraulic Enclosure per their design.
There were no safety consequences impacting the plant or public safety as a result of this event. All control rods fully inserted and plant response was as expected. This event is being reported pursuant to 10CFR50.73(a)(2)(iv)(A) due to an actuation of the Reactor Protection System (RPS).
|05000219/LER-2013-003||16 January 2014||Oyster Creek|
On 11/17/13 at 1725 EST station personnel simultaneously opened the inner and outer air lock doors from the Reactor Building to the reactor control enclosure (Reactor Building 23 foot elevation North West airlock). Reactor Building secondary containment integrity was momentarily declared inoperable per Technical Specification (TS) 3.5.B.2 due to both containment airlock doors being simultaneously open at the same time. Reactor Building D/P (differential pressure) remained less than negative 0.25 inches water column. The doors were immediately closed within 10 seconds and the Technical Specification condition was exited. The cause was a degraded airlock door interlock.
The secondary containment interlock doors were open for less than ten seconds. Based upon the short duration of the secondary containment doors being opened simultaneously, individuals present in the airlock to close the doors and that the Reactor Building D/P remained less than negative 0.25 inches water column during the course of this event, the safety function was met and this event is of low safety significance.
This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(C), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.
|05000219/LER-2013-002||5 December 2013||Oyster Creek|
On October 6, 2013 at approximately 1040 EDT, during a planned reactor power ascension with reactor power at approximately 20% of rated thermal power, main condenser vacuum began to lower. In accordance with the abnormal operating procedure for degrading vacuum, Operators inserted a manual scram of the reactor at 1130 EDT.
Following the reactor SCRAM, operations and maintenance personnel identified an approximate 1" hole on Y-1-26, 'B' Condenser Steam Inlet Expansion Joint on the south side of 'B' Condenser. It was confirmed to be an active leak and subsequently the source of condenser vacuum degradation. A temporary leak repair was performed.
All control rods fully inserted and plant response was as expected. This event is being reported pursuant to 10CFR50.73(a)(2)(iv)(A) due to an actuation of the Reactor Protection System (RPS).
|05000219/LER-2009-005||5 February 2010||Oyster Creek|
On July 12, 2009 with the Unit at 100% power in the "Power Operation" mode, a severe electrical storm resulted in multiple lightning strikes on an 'interconnected 34.5 kV offsite transmission line. These lightning strikes in conjunction with a failure of a line breaker to open caused grid disturbances, a main generator trip on over-excitation, an automatic reactor scram due to the load rejection, and a loss of offsite power to the Startup Transformers.
When the plant experienced the loss of offsite power to the Startup Transformers, both Emergency Diesel Generators (EDGs) started and energized their associated safety related buses. During the event, EDG #1 was slow to tie onto its safety-related bus and the erratic "B" Isolation Condenser (IC) level indication resulted in operators declaring the IC inoperable. An Unusual Event was declared for the loss of power to the Startup Transformers for greater than 15 minutes. All declarations and notifications were made correctly and in a timely manner.
This event is being reported pursuant to: 10CFR50.73(a)(2)(iv)(A) due to automatic actuation of the reactor protection system, isolation condensers, and emergency diesel generators; and 10CFR50.73(a)(2)(v)(D) due to loss of offsite power to the Startup Transformers.
|05000219/LER-2009-004||12 November 2009||Oyster Creek|
During the 1R22 Refueling Outage (Fall 2008), a Temporary Modification was made to the secondary containment so that the trunnion room door could be left open to support maintenance activities. The Temporary Modification moved the barrier for secondary containment from the outer walls and door of the trunnion room to the inner walls, ceiling, floor, and associated penetrations in the trunnion room. A 10 CFR 50.59 Evaluation supported the Temporary Modification based on the determination that secondary containment integrity could be maintained separate from the trunnion room door and concluded that a Technical Specification change was not needed.
During the NRC Plant:Modification Inspection (May 4 through 15, 2009), the determination was made that the Temporary Modification. to open the trunnion room door for extended periods of time during 1 R22 was in noncompliance with Oyster Creek's Technical Specifications. The supporting 10 CFR 50.59 Evaluation inappropriately determined that prior approval from the NRC was not needed in that the secondary containment integrity required by Technical Specification Limiting Condition of Operation (TS LCO) 3.5.B was maintained.
|05000219/LER-2007-001||17 September 2007||Oyster Creek|
On July 17, 2007 at 05:21 while operating at 100% power, an automatic reactor scram occurred due to low reactor water level following a trip of the "C" Reactor Feed Pump (RFP). The cause of the "C" RFP trip is attributed to an electrical fault internal to the motor. This transient led to an automatic scram on
There were no safety consequences impacting plant or public safety as a result of this event. .
This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to the automatic reactor protection system and subsequent ECCS actuations.
|05000219/LER-2002-003||2 April 2004||Oyster Creek|
Apparently, sand may not have completely filled the area and/or had settled over time creating this void.
The safety significance of this discovery is minimal, as there Is no combustible material in the void. Both cables are contained in conduit and have sufficient Class 1 E electrical separation.
Immediately upon discovery, a continuous fire watch was stationed. Additional actions were subsequently taken to open communication between the void and adjacent smoke detectors. This would provide early warning early warning of a degraded condition. Adequate separation will be provided by re-filling the void between the two 4160 VAC feeder conduits to meet Appendix R requirements. The sand fill will be monitored. No similar events have occurred.