|Report date||Site||Event description|
|05000287/LER-2017-001||20 September 2017||Oconee|
Oconee Nuclear Station Unit 3
On 7/24/17, with Oconee Nuclear Station (ONS) Unit 3 operating at 100 percent power, Transmission Department Relay personnel were in the ONS 525kV Switchyard Relay House performing preventive maintenance on a Breaker Failure Relaying device for Power Circuit Breaker PCB-57. This is a non- safety PCB that isolates a commercial transmission line from the commercial bus in the 525kV switchyard.
The maintenance was intended to actuate the protective relaying for PCB-57. The crew inadvertently connected test equipment to the adjacent relaying for PCB-58. The activation of the PCB-58 relay resulted in a Unit 3 separation from the electrical grid and a generator "Lockout." The lockout generates a turbine trip which in turn trips the reactor via the Reactor Protection System (RPS). This actuation of the RPS is reportable per 10 CFR 50.73(a)(2)(iv)(A).
Post trip plant response was normal and plant conditions were controlled and maintained within the allowances of Technical Specifications with no personnel injuries or safety system actuations.
A cause analysis attributed the cause of this event to human error in that test equipment was inadvertently connected to relaying for the incorrect PCB. The cause analysis corrective actions will address the likelihood of comparable human errors from occurring.
|05000287/LER-2013-001||3 June 2016||Oconee|
This revision provides supplemental information, including a revised cause, discovered during the cause evaluation for a similar Unit 3 manual reactor trip on January 31, 2015.
On October 24, 2013, Oconee Unit 3 was at 100% power when Operators observed abnormal Main Feedwater flow oscillations. Manual control was unsuccessful in stabilizing the oscillations. The Control Room supervisor directed a manual trip at 0553 hours. Four main steam relief valves (MSRV) did not fully reseat. Post-trip procedures were used to reseat the MSRVs. All other post trip conditions were normal. The 2013 cause evaluation identified a failure of a bushing seal (o-ring) in the actuator for Main Feedwater Control Valve (MFCV) 3FDW-32 as the cause of the flow oscillations. The 2015 cause evaluation revealed the cause of the 3FDW-32 control problem was a latent defect which produced an intermittent fault in the voltage to pneumatic (E/P) converter. While the o-ring failure was a plausible cause in 2013, given new information from the 2015 event, the faulted E/P converter was determined to be the cause of the 2013 event. The E/P converter was tested in 2013, but due to the intermittent nature of the fault the E/P was refuted after satisfactory testing of 3FDW-32 controls.
This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as a manual actuation of the Reactor Protection System (RPS).
|05000269/LER-2016-001||5 May 2016||Oconee|
At 1512 on March 6, 2016, Oconee Nuclear Station (ONS) Unit 1 experienced a Reactor Trip, initiated by an electrical bushing failure on the Unit 1 Main Step-up Transformer that resulted in a transformer fire. The plant response to the trip was evaluated, and was found to be acceptable. The reactor trip is a RPS actuation and requires an LER per 10 CFR 50.73(a)(2)(iv)(A).
Due to the transformer fire a "Notification Of Unusual Event" was declared at 1520. The fire eventually led to a transformer overhead line failure that caused a switchyard bus lockout and the loss of one emergency power path for the site. The second emergency power path and multiple offsite sources remained operable during the event. The emergency classification was upgraded to an "Alert" condition at 1658, upon determining that the fire had an impact on safety related equipment.
The 4-hour notification related to this reactor trip was bounded by the notifications made for the Emergency Classifications described above (See Event Notification 51770).
Post trip conditions and Emergency event conditions were controlled and maintained within the allowances of Technical Specifications with no personnel injuries or challenges to other safety system actuations. The transformer deluge system activation along with the response of onsite and offsite fire teams promptly contained the fire.
|05000287/LER-2015-001||31 March 2015||Oconee|
On January 31, 2015, Oconee Unit 3 was operating at 100% power in MODE 1 when Control Room operators observed that Main Feedwater flow indicators were oscillating outside of normal parameters. The Control Room supervisor made the decision to manually trip Unit 3 at 1431 hours due to erratic feedwater operation and increasing RCS pressure. A subsequent investigation determined the feedwater flow oscillations were caused by a subcomponent failure of the electrical to pneumatic converter (E/P), 3FDW EP 0007, to properly control feedwater flow for Main Feedwater Control Valve (MFCV) 3FDW-32.
This event was reported as a 4-hour notification to the NRC on January 31, 2015, in Event Notification (EN) number 50781 under 10 CFR 50.72(b)(2)(iv)(B) - Reactor Protection System (RPS) Actuation - Critical. The event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an actuation of the RPS.
|05000287/LER-2015-001, Manual Reactor Trip Due to Unacceptable Main Feedwater Flow Control Valve Oscillations||31 March 2015||Oconee|
|05000269/LER-2014-002||26 January 2015||Oconee|
On November 25, 2014, Duke Energy Carolinas, LLC (Duke Energy) reviewed AREVA 10 CFR 50.46 Notification Letter FAB 14-00631, which indicated a deficiency had been discovered in the uranium thermal conductivity models used in the Oconee Nuclear Station (ONS) Loss of Coolant Accident (LOCA) analysis of record. When the model deficiency was corrected, the ONS Large Break LOCA Peak Cladding Temperature (PCT) exceeded 2200 degrees F. The Oconee licensing basis PCT is evaluated for compliance with the criterion in 10 CFR 50.46(b)(1) and must not exceed a PCT of 2200 degrees F. This issue was reported to the NRC as an 8-hour non-emergency notification on November 25, 2014 (NRC Event Notification EN 50640). On December 17, 2014, Duke Energy submitted a Special Report as required by 10 CFR 50.46 (ML14353A214).
Once notified, Duke Energy implemented administrative restrictions to ensure that the Oconee Units would not violate the reduced linear heat rate (LHR) limits specified by AREVA.
There are no new regulatory commitments contained within this Licensee Event Report.
|05000269/LER-2014-001||27 May 2014||Oconee|
For the 13.8kV Keowee Hydroelectric Station (KHS) power cables contained in the underground trench to both transformer CT-4 and to the Protected Service Water (PSW) system, single failure analysis considered a three-phase bolted fault at the end connections and a phase-to-ground fault as bounding conditions. This single failure analysis was questioned in a recent NRC inspection as not being bounding. Duke Energy evaluated the issue and concludes the cables to be operable but nonconforming with the current licensing basis due to a lack of documentation. Additional cable analysis and/or testing is needed. The results of planned power cable analysis and/or testing will provide the necessary information needed to update the licensing basis documentation. This condition is conservatively being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B) as an event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
The preliminary apparent cause of this event is a missed opportunity to update the licensing basis in 2002 when the underground trench and cables were modified, and again in 2013, when the PSW system 13.8kV power cables and control cables were placed in service. Duke will complete the causal evaluation and as necessary, supplemented the report following the conclusion of that evaluation. Duke Energy used a risk-informed approach to determine the risk significance associated with the condition. The analysis results concluded that the postulated single failure of the Keowee emergency power supply represents an insignificant impact to plant risk.
|05000269/LER-2013-002||26 August 2013||Oconee|
|05000269/LER-2013-001||8 April 2013||Oconee|
On 2/6/2013, with all three Oconee Units at 100 percent power, Oconee Nuclear Station (ONS) personnel concluded that emergency power equipment could be adversely impacted by a licensee identified original design issue involving inadequate analysis of electrical equipment heat loads and weaknesses in the Heating Ventilation & Air Conditioning (HVAC) system design.
The investigation determined that the principle causes were associated with inadequate and incomplete inputs and methods in development of HVAC systems during original plant design and Blockhouse, and the 230kV Switchyard Relay House susceptible to single failures.
Compensatory measures were put into place. Corrective actions include modifications to resolve the inadequacies with the original plant design issues.
This event is reportable under the following criteria: Operations Prohibited by Technical Specifications, Unanalyzed Condition, Event that could have prevented Fulfillment of a Safety Function, and Common Cause.
|05000287/LER-2009-001||10 February 2009||Oconee|
On December 12,T 2008 two containment isolation valves in a Post Accident Liquid Sampling line in Unit 3 were declared inoperable.T Technical Specification Limiting Condition for OperationT (LCO)T 3.6.3 Conditions A and B were entered immediately and power was removed from two normally-closed solenoid-operated valves in the same line to meet the required actions of the LCO.T However,T the unit has been operated since the valves were installed in 1996 until discovery of the inoperable. condition.'T Consequently,T the appropriate Technical Specification required actions were not met within the allowed outage time permitted, and the unit was operated in a condition prohibited by Technical Specifications.
The inoperable condition was caused by the use of soft seat materials which are not qualified for the elevated pressure and temperatures they could be exposed to in post-accident sampling service.T The planned corrective action is to replace the soft-seated valves with a hard-seated design.
This event is considered to have no significance with respect to the health . and safety of the public.