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On May 31, 2016 at 1626, Calvert Cliffs Unit 1 experienced an automatic reactor trip. The cause of the trip was a spurious high level steam generator signal due to a failed Steam Generator 11 Channel B High Level Turbine Trip Under Voltage Logic Module, 12/4BL-XA27.

The spurious high level trip signal resulted in a turbine trip followed by an automatic reactor protection system trip on loss of load. The failed under voltage logic module was sent off-site to the vendor to conduct a failure analysis. The analysis identified two most probable causes for the failure. The first most probable cause was due to an intermittent failure of an integrated circuit of the under voltage logic module. The second most probable cause was due to an inadvertent solder bridge between a pin on the integrated circuit chip and the board. A refurbished logic module was installed and subsequently tested prior to the unit returning to power. The corrective action planned to prevent recurrence is to replace the current Engineered Safety Features Actuation System (which includes the non-safety high level turbine trip) with a system that will eliminate single point vulnerabilities within this system.

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On February 20, 2016, during Calvert Cliffs Inservice Inspections of Unit 1 dissimilar metal welds, an evaluation of recorded ultrasonic examination (UT) data identified the presence of one axially oriented flaw in a 4 inch Unit 1 pressurizer (PZR) safety relief nozzle (NZL) to safe end weld. The indicated flaw exhibited characteristics indicative of primary water stress corrosion cracking (PWSCC). The flaw was inner diameter connected and measured 81.6 percent through-wall. This measured axial flaw depth did not meet the American Society of Mechanical Engineers (ASME) Code allowable limit. The UT of this weld done in 2006 and 2010 had determined the axial flaw depth was only approximately 8 percent through-wall. After re-analysis of the prior UT data, the root cause was determined to be limitations in sizing data collection and analysis techniques prior to 2016 that were unable to connect the detected inner .

diameter indication to the full through-wall extent of ultrasonic signal response of the flaw. The weld was repaired using a full structural weld overlay repair method using PWSCC resistant material deposited around the circumference of the weld area. Two other Unit 1 welds that had previously shown potential PWSCC were examined with the new techniques and showed no change in their indication characteristics and remain within ASME Code allowable limits.

During scheduled testing at the offsite testing facility, the as-found lift settings for the pressurizer safety valves (PSVs) previously installed in Unit 1 at the 1RV200 and 1RV201 locations were measured outside the Technical Specification allowable values (both valves lifted low). The valves had been installed during the 2012 Unit 1 refueling outage and were removed during the 2014 Unit 1 refueling outage for scheduled testing and maintenance.

Spare valves were installed during the 2014 refueling outage. The failed valves were disassembled and inspected at the offsite facility. The apparent cause of the PSV failures is that the internal lift spring assemblies of a specific manufacturer lot failed to hold PSV set pressure. This apparent cause will be verified as the removed internal lift spring assemblies for each valve is further reviewed, examined, and tested to support this apparent cause. No other installed PSVs contain internal lift spring assemblies from this specific manufacturer's lot.

On January 21, 2014, at 9:25 p.m., Unit 2 experienced an automatic reactor trip from 99.5 percent power. The reactor trip occurred when 13 kV Service Bus 21 deenergized due to a ground fault on feeder breaker 252-2104. The loss of the service bus caused a loss of power to non-safety-related 4 kV buses, which caused a loss of circulating water pumps and main condenser vacuum, requiring the use of auxiliary feedwater and atmospheric dump valves to maintain Reactor Coolant System temperature. Power was also lost to safety-related 4 kV Bus 24, which caused an automatic start of the 2B Emergency Diesel Generator to power 4 kV Bus 24.

The loss of power to non-safety-related 13 kV Service Bus 21 was caused by water intrusion when an air filter assembly located at the back of breaker 252-2104 cubicle became dislodged during a snow storm, allowing snow to enter the cubicle, melt, and cause a ground fault. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to Reactor Protective System actuation and the automatic start of the 2B Emergency Diesel Generator. Corrective actions include repairs to 13 kV Service Bus 21 and installation of a new filter housing. Previous events related to safety-related structures have been documented in Licensee Event Reports 317/2010-001 and 317/2011-003.

On October 21, 2011, Calvert Cliffs Nuclear Power Plant discovered that a reportable condition existed. On August 28, 2011, numerous alarms were received for the 1A Emergency Diesel Generator (EDG) (train A, one of two safety-related EDGs dedicated to Unit 1). Water was intruding down the diesel generator intake piping, resulting in a short circuit on the 1A2 engine speed switch circuit. The 1A EDG was declared inoperable and appropriate Technical Specifications were implemented. The 1A2 engine speed switch assembly was cleaned, dried, and inspected. The 1A EDG was returned to operable status. Corrective actions include repair of the affected piping penetrations.

A root cause analysis determined the most probable root cause to be the penetration around the 1A2 engine combustion air intake pipe on the 80 foot level of the 1A EDG Building was unable to perform its design function of being leak tight. This may have occurred due to improper installation, manufacturing defect, degradation, or original design considerations. This condition may have existed since construction of the 1A EDG Building in 1996. There have been no licensee event reports for similar events at Calvert Cliffs.

On August 27, 2011, at 2248 eastern daylight time, Unit 1 experienced an automatic reactor trip from 100 percent power. The Reactor Protective System actuated on loss of load. The loss of load occurred due to a phase-to-phase short circuit on the main transformer when main transformer lines were struck by dislodged Turbine Building siding caused by winds associated with Hurricane Irene. Immediately following the short circuit, 14 Containment Air Cooler stopped operating. Shortly after the plant trip occurred, 1A Emergency Diesel Generator was declared inoperable due to a shorted speed switch. The root cause analysis performed to address this event concluded that the Turbine Building Northwest corner siding was not installed per design during original construction. This resulted in a weaker siding connection to the Turbine Building structure, allowing the siding to come off in wind speeds less than design.

At Calvert Cliffs, there have been no recent similar events involving a reactor trip associated with severe weather. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to Reactor ProteCtive System actuation. Corrective actions include testing and inspection of the main transformer, replacement of B and C phase high line drops to the main transformer and inspection and repair of electrical connectors on the 1A Emergency Diesel Generator.

On May 12, 2010 at 13:51, Unit 1 experienced an automatic reactor trip from 100 percent power, when the Reactor Protective System actuated on high pressurizer pressure due to a complete load rejection. At the time of the event, maintenance was being performed in the switchyard requiring two 500 kV breakers to be open with a third 500 kV breaker (552-22) closed, thus allowing Unit 1 main turbine generator output to the grid. However, due to a loose connection in the 125 VDC distribution panel board for the breaker disconnect switches, breaker 552-22 tripped open upon receiving a trip signal from its trip circuitry, resulting in the complete load rejection condition. The root cause of the event was a loose electrical connection in the 125 VDC distribution panel board. After the trip, one of the pressurizer safety valves leaked. The Unit was taken to Mode 5 to repair the leaking pressurizer safety valve.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to a valid actuation of the Reactor Protective System. Corrective actions include checking the electrical connections in the 125 VDC distribution panel for tightness, increasing the time delay to the re-trip feature of the subject type relays, replacing the disconnect bucket and revising the risk management procedure.

On December 17, 2006, at 04:25 a.m. plant personnel determined that the Reactor Protective System (RPS) Rate of Change of Power-High (or Startup Rate (SUR)) Trip would not be enabled at the Technical Specification required value of 12 percent Rated Thermal Power (RTP), decreasing, on Unit 1 RPS Channel A and Unit 2 RPS Channel C. This condition was identified RPS Channel A arid Unit 2 RPS Channel C recorded the Nuclear Instrument (NI) Level 1 Bistable resets as 11.9 percent and 11 percent RTP by local indication, respectively. The failure of the SUR Trip enabling function to reset prior to 12 percent RTP decreasing is prohibited by Technical Specifications. This condition rendered the automatic bypass removal feature inoperable for the affected channels. This event was caused by a failure to recognize that when Improved Technical Specifications were adopted at Calvert Cliffs Nuclear Power Plant in 1997, new acceptance criteria were added to the Technical Specifications for this function. The affected Unit 1 and Unit 2 SUR Channel Trip Units were placed in bypass mode upon discovery of the condition. Corrective actions to ensure the automatic bypass removal setpoints were within specifications for Unit 1 and 2 were completed at 11:30 p.m. on December 17, 2006.

Technical Specification Bases and STP changes will be made to prevent future recurrence.

On March 24, 2006 with Unit 1 shutdown in Mode 6, Motor Control Center (1MCC123) Feeder Breaker (52-1703) tripped on short-time overcurrent during performance of a surveillance test procedure. Unit 2 was operating in Mode 1 at the time of the event. The feeder breaker powers safety-related auxiliaries required for operation of the 1A Emergency Diesel Generator (EDG).

As a result of the feeder breaker tripping, operations personnel secured the 1 A EDG and secured the surveillance test. Subsequent investigation determined that the short-time overcurrent amptector setpoint was set too low and had drifted down to a value lower than the associated inrush starting current, resulting in the feeder breaker trip. An engineering evaluation determined that the original amptector design setpoint, established by a vendor, was not adequate because the setpoint did not consider all potential loads that could be realized upon an EDG start and load during a design basis event. Corrective actions included installing a new amptector, increasing the amptector setting, and notifying the vendor. Additional actions planned include training engineering personnel and evaluating safety-related feeder breakers to ensure setpoints have adequate margin. The 1A EDG was subsequently tested satisfactorily and returned to service on March 25, 2006. This event is also reportable per 10 CFR Part 21.