Semantic search

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).

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.

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.

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.

• This event resulted in an Operation or Condition that was Prohibited by the Plant's Technical Specifications (TS) and is therefore being reported under 10CFR50.73(a)(2)(i)(B).

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

• Steam Isolation Valve Room (Trunnion Room). The purpose of the TCC is to Isolate the penetration from the Reactor Building (RB) to the Trunnion Room to allow the RB Trunnion Room door to be maintained open during refueling outages. Per Technical Specifications (TS) 3.5.8 and 4.5.G, in order to maintain secondary containment integrity with the Trunnion Room door open, four -1 penetrations connecting the Trunnion Room to the RB must be seated. The Trunnion Room door was open when the phi() was found removed from a 6" equipment drain hub. The plug was immediately reinstalled. On September 20, 2014, the Post Maintenance Testing (PMT) for the TCC reinstallation was executed in order to declare secondary containment operable. This PMT revealed that the Standby Gas Treatment System (SBGTS) could not achieve the required minimum differential pressure of - 0.26° water vacuum on the RB. The cause was determined to be an improperly latched personnel access hatch on'the Outer Reactor Building Airlock Door.

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.

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).

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).

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).

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.

• A void was discovered in an area expected to be filled with sand beneath a portion of the two 480 VAC switchgear rooms. This void created an open area between two 4160 VAC feeder conduits. Because of the void, Appendix R electrical separation criteria were no longer met.

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.

1