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On 10/23/2017, during a walkdown of containment at the start of the Unit 3 refueling outage, a leak was identified in a socket weld for a 1-inch diameter instrument line. The line is connected to discharge piping for the 'B' recirculation pump and is part of the reactor coolant system pressure boundary. Because the leak was misting, the leakage rate could not be quantified. However, the reactor coolant system unidentified leakage prior to plant shutdown was 0.18 gpm. RCS pressure boundary leakage while in Mode 1 is a violation of Technical Specification 3.4.4 and is a reportable condition.

The cause of the event was a lack of fusion defect in the weld when it was done in the late 1980's. Normal vibration of the line since it was installed resulted in the crack initiating at the weld defect and propagating to the surface. The section of pipe and associated fitting were replaced, along with welds in similar sections of piping. There were no actual safety consequences as a result of this event.

On 12/31/15, at approximately 0630 hours, during shift turnover panel walk-downs, a licensed operator identified that the High Pressure Coolant Injection (HPCI) flow controller output indication was showing zero percent. The controller was in automatic with the set point at 5000 gpm, which would typically result in a controller output value of 100%.

This condition would have prevented the HPCI system from performing its design function in the event of an accident.

The HPCI system was declared inoperable and the appropriate Technical Specification Action was entered.

Other standby systems (Reactor Core Isolation Cooling and low pressure emergency core cooling systems) remained operable.

Troubleshooting of the flow controller and related circuitry identified a failed signal converter.

The signal converter was replaced and the HPCI system was declared operable at approximately 2110 hours on 12/31/15.

There were no actual safety consequences associated with this event.

On October 4, 2012, during surveillance testing of a 4 kV emergency bus, Operations personnel were in the process of placing the Main Control Room Emergency Ventilation (MCREV) System in service. When an initiation signal was generated, with the "K MCREV fan control switch positioned in 'Auto', the fan did not start as expected. The "B" MCREV fan started approximately 45 seconds later, as designed for the fan in standby. The bus that powers the "B" MCREV train was energized, but was considered inoperable due to testing and other work being performed. With both MCREV subsystems inoperable, Technical Specification (TS) 3.7.4 Condition E was entered, which requires the plant to be in Mode 3 in 12 hours.

The cause of this event was that the switch was positioned such that the switch contacts were still open, which prevented a start of the "A" MCREV fan. The switch position indication is not precise enough to ensure the switch contact block is set for a given indicated position.

The switch was repositioned, normal system lineup was restored and the system was returned to an operable status. This event has been placed in the station's corrective action program to further evaluate and track additional actions.

There were no actual safety consequences as a result of this event.

reactor from 0.2% power as a result of an observed shortening reactor period. The reactor was manually scrammed by placing the mode switch in the shutdown position. There were no actual safety consequences associated with this event. The scram function operated as designed and there were no complications as a result of the manual scram. The event was considered not to be risk significant.

The cause of the manual scram was due to a shortening reactor period during a planned soft shutdown using control rods. The manual scram was required due to a small amount of feedwater being added to the reactor during a time period when control rods were not being inserted. This resulted in a decrease in the reactor period. Plant shutdown procedures for Units 2 and 3 will be revised to provide additional control of operational activities to minimize the likelihood of a short period when performing a soft shutdown.

There were no previous similar LERs identified.

Based on information received on 10/2/05 from a laboratory performing Safety Relief Valve (SRV) as-found testing, Site Engineering personnel determined that SRV set point and performance deficiencies existed with five SRVs that were installed during the 15th operating cycle for Unit 3. Four of the SRVs were determined to have their as-found set points in excess of the Technical Specification allowable ± 1% tolerance. In addition, one additional SRV was found to not properly re-close when tested. The cause of the four SRVs being outside of their allowable as-found set points is due to set point drift. Concerning the failure of the SRV to re-close, the preliminary laboratory failure analysis identified that the main valve disc had not properly re-seated when closing due to misalignment of the main valve disc spring. The valve was last refurbished in February 2001.

The five SRVs were replaced with different SRVs for the 16th Unit 3 operating cycle. Additional assessment and appropriate corrective actions concerning SRV refurbishment vendor work practices will be further assessed as part of the Corrective Action Program. There were no actual safety consequences associated with this event.

NRC FORM 366 (6-2004) PRINTED ON RECYCLED PAPER NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001) FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)

On 9/12/16, an engineering evaluation concluded that a flaw identified on a 1" diameter stainless steel pipe associated with the High Pressure Service Water (HPSW) System could have failed during a seismic event.

Failure of the pipe would cause flooding of the '2C' Residual Heat Removal (RHR) pump room and impact operation of equipment in the room if needed during a design basis event. The 1" diameter pipe supplies sample water to the '2C' HPSW radiation monitor sample pump. The flaw was identified on 8/16/16 when a 120 drop per minute leak was observed. The leak was the result of a crack located in the pipe at the toe of a fillet weld connecting the pipe to a coupling on an 18" diameter pipe that returns HPSW water to the plant discharge canal.

There were no actual consequences as a result of the leak. The pipe was replaced and the equipment was returned to service.

On 10/29/14 at approximately 1100 hours, Engineering personnel determined that the N-9A primary containment penetration pathway had leakage that exceeded the maximum allowable primary containment leakage rate (La) value required by Technical Specification (TS) 5.5.12, Primary Containment Leakage Rate Testing Program. This determination was based on as-found leakage through the seats of two redundant feed water check valves (CHK-2-06-28A and CHK-2-06-96A) that are Primary Containment Isolation Valves (PCIVs).

The cause of the deficiency in both swing check valves was due to operational wear on the check valve pivot shaft and associated bushings. Both check valves were repaired during the Refueling Outage and as-left leak testing proved appropriate leak-tightness of the check valves in the closed position.

Additional corrective actions are being assessed as part of a causal analysis being performed in accordance with the station's corrective action program. There were no actual safety consequences associated with this event.

On May 21, 2014 at approximately 1100 hours, based on inspections being performed as part of an extent- of-condition review, it was determined that an unanalyzed condition existed that potentially impacted the ability to mitigate an Appendix R fire postulated to occur in the control room and cable spreading room.

Broken wires leading to the alternate fuse for the 125vdc control power supply had been previously identified in the breaker enclosures for three 4 kV safety-related breakers. In the event of an Appendix R fire in the control room and cable spreading room, with a fire-induced short circuit that results in a blown primary control power fuse, the broken wires would result in the loss of control power to the affected breakers. This would impact the ability to close the breaker locally after the control room has been evacuated. Instead of using the handswitch on the front of the breaker enclosure to close the breaker, the operator would have to open the enclosure door and push the manual close button on the breaker.

Based on the plant's fire protection defense-in-depth features and the ability of the operator to perform a simple alternate action to close the breaker, this event would not impact the ability to reach and maintain a safe shutdown condition in the event of an actual fire.

On 10/7/06 at 1802 hours, an Unusual Event was declared for Unit 2 as a result of the discovery of a leak at an elbow for piping that penetrates the Primary Containment Suppression Pool (i.e. Torus). The 4" piping is the High Pressure Coolant Injection (HPCI) / Reactor Core Isolation Cooling (RCIC) Torus Flush line. This line is normally isolated from the HPCI / RCIC systems by a closed motor-operated valve and is only used during testing activities. The leak was discovered by an equipment operator at approximately 1741 hours during a planned inspection associated with a RCIC system check valve. The leak occurred on the intrados of a 45 degree elbow of the 4" piping. The elbow was located approximately 1 foot above the Torus penetration (i.e.

the leak was outside of Primary Containment). The cause of the crack in the elbow is due to cavitation and abrasive erosion and/or localized water-jet cutting resulting from excessively high flow velocities through this piping during test conditions in conjunction with an apparent lack of fusion between the weld backing ring and the weld root at the elbow weld. The leaking elbow was replaced and non-destructive testing was performed.

The similar pipe on Unit 3 was examined and no significant concerns were noted. Extensive walk downs of similar piping that is attached to the Torus was conducted for both Units 2 and 3. There were no similar deficiencies discovered. Selected ultrasonic testing was performed on Unit 2 and 3 Torus attached piping that involved higher flow rates. These examinations also did not identify any similar concerns.