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APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 Reported lessons learned are incorporated into the licensing process and fed back to industry.

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On February 10 through February 13, 2017, planned operations with the potential to drain the reactor vessel (OPDRV) activities were performed while Unit 2 remained in Refueling Mode (Mode 5) without an operable secondary containment. These activities are prohibited by Technical Specifications (TS) 3.6.4.1. However, recent NRC guidance provided in Enforcement Guidance Memorandum (EGM) 11-003, Revision 3 and in Regulatory Issue Summary (RIS) 2012-11 allowed the implementation of interim actions as an alternative to full compliance. These actions are contained in the operating procedure for the OPDRV activities performed during the 2017 Hatch Unit 2 Refueling Outage in February.

The performance of OPDRV activities were planned activities that were completed under the guidance of plant procedures and confirmed to have low safety significance based on interim actions taken. Since these actions were deliberate, no cause determination was necessary. A license amendment request (LAR) will be submitted for the associated NRC-approved Technical Specifications Task Force (TSTF) Traveler 542, "Reactor Pressure Vessel Water Inventory Control.

On August 18, 2016 at 1006, while performing the monthly Emergency Diesel Generator (EDG) surveillance, the 2C Limiting Condition of Operation (LCO) 3.8.1.b. The other four EDGs were successfully started for surveillance testing in accordance with Tech Specs to preclude a common cause failure mode. Further inspections were performed and the cross drive shaft between the pump flexible drive gear and the engine driven lube oil pump was found sheared.

The cross drive shaft assembly was replaced and the 2C EDG was declared operable on August 24, 2016 after subsequent functional and post maintenance testing.

Investigations into the failure revealed that the initial cross drive shaft fracture existed prior to the monthly surveillance performed on August 18. The time frame for which the 2C EDG was able to perform its safety function for its 7-day mission time is indeterminable due to the failure mechanism of the cross drive shaft. A loss of safety function is being reported due to the 2C EDG being potentially inoperable during a time in which either the 1B swing EDG was inhibited from Unit 2 or the 1B/2A EDG was inoperable due to normal equipment maintenance and testing.

Additionally, this report constitutes a Part 21 notification per 10 CFR 21.21(d).

On 03/17/2016, with Unit 2 at 100 percent rated thermal power (RIP), a fuel oil leak was observed at the inlet to the fuel oil relief valve during the performance of the 2C Emergency Diesel Generator (EDG) semi-annual surveillance. The 2C EDG was declared inoperable until the relief valve was replaced. Investigation revealed that the leak was due to a through wall indication in the threaded area of the relief valve. Inspection indicated that the failure was consistent with flaws observed in cases of fatigue failures. Site analysis of the failure provided reasonable assurance of historic operability; however a vendor was engaged to perform an independent review with a detailed seismic model of the system.

On 4/13/2016, the vendor analysis indicated that a seismic event could cause the degraded fuel oil relief valve to adversely affect EDG operation. The analysis concluded that although the 2C EDG could have met its 7-day mission time with the existing leakage rate, a design basis earthquake could reduce the ability of the diesel to maintain the minimum fuel header pressure and sufficient amount of fuel for 7 days. Thus the seismic qualification of the 2C EDG was degraded, but the safety function of the emergency AC system would only be challenged under certain circumstances with a seismic event.

Corrective actions will establish periodic replacement of similar fuel oil relief valves on the EDGs.

On September 13, 2013 at 1438 EDT, with the unit in Mode 1 at 100 percent power, the Unit 2 'D' outboard main steam isolation valve (MSIV) failed to move during surveillance testing that involved partial closure testing of the valve. Based on previous operating experience this test result typically indicated a problem with a solenoid used for partial valve closure testing and was not indicative of a problem with the MSIV. A plan was made to cycle the MSIV to the "closed" position and return it to the "open" position during a planned power reduction on September 14, 2013. If the failure was indeed due to the test solenoid, the affected MSIV would function as expected after the power reduction.

At 1105 EDT on September 14, 2013, following power reduction to approximately 65 percent power, the 'D' outboard MSIV failed to close using its control switch. Actions were taken to isolate the MSIV penetration as required by Technical Specifications. Maintenance activities were completed and the MSIV underwent rigorous testing with successful results to confirm the operability of the 'D' outboard MSIV.

On March 18, 2013 at 0910 EDT, the Unit 2 High Pressure Coolant Injection (HPCI) system was declared inoperable after the Turbine Control Valve (TCV) did not open as expected during the HPCI Pump Operability 165 psig Test. Unit 2 was in Mode 2 (Startup) and was not in power production at the time of the event. The TCV failed to open because the hydraulic lines between the hydraulic actuator and the remote servo had been interchanged during performance of the 10-year preventive maintenance (PM) procedure in the 2013 refueling outage for Unit 2.

The lines were properly reconnected, and the HPCI system returned to Operations at approximately 1033 EDT on 3/19/2013. The HPCI operability surveillance was successfully performed at approximately 1056 EDT on 3/19/2013. Additionally, the PM Procedure was revised to include more detailed instructions on the labeling of the hydraulic actuator and remote servo piping prior to removal and to include additional post-maintenance testing criteria for the hydraulic actuator and remote servo.

On 2/16/2013, at approximately 0310 eastern standard time (EST), Unit 2 was in the refueling mode, when a valid unplanned reactor protection system (RPS) actuation occurred as a result of scram discharge volume (SDV) high level. Following shutdown for refueling the scram vent and drain valves would not open as expected which resulted in water remaining in the SDV. The refueling interlocks surveillance testing was being performed in accordance with the Operations surveillance procedure with the SDV water level above the setpoints for the "rod withdrawal block" and the "scram trip" in conjunction with closed SDV vent and drain valves. Jumpers had been installed to bypass the rod block signal, and the scram signal had been bypassed using the "Disch Vol Hi Level Byp" switch.

When the Reactor Mode Switch was placed in the "Start & Hot Standby" position, a full reactor protection system (RPS) actuation occurred. Subsequent investigation revealed that the SDV high level bypass is only active with the mode switch in the "Shutdown" or "Refuel" positions.

The RPS actuation was caused by the presence of a water level in the SDV above the trip setpoint and by a less than adequate procedure. The absence of procedure prerequisites to confirm the SDV Hi Level Rod Block and RPS trip signals were not present prior to performing the refueling interlocks surveillance procedure resulted in positioning the reactor mode switch to the "STARTUP" position with the SDV trip signal present. This caused the SDV high level trip signal to no longer be bypassed resulting in the RPS actuation. The SDV high level condition was cleared and beginning of shift training (BOST) was issued to inform licensed operators of the event. Actions to prevent recurrence involved revision of Units 1 and 2 refueling interlocks surveillance procedures.

On October 24, 2011, at approximately 0% power during startup from a scheduled maintenance outage, the 'A' IRM signal showed increasing levels of electrical noise while on Range 1. A spike in the signal resulted in a half-scram signal that prompted Operations personnel to bypass the 'A' IRM and declare it inoperable. The 'C' IRM subsequently began exhibiting erratic behavior and slowly drifted downscale while on Range 7. Operations personnel "ranged" down the 'C' IRM. Its signal continued to display the same behavior. Operations personnel declared the 'C' IRM inoperable, resulting in no operable IRM channel in one quadrant of the reactor core. Further control rod withdrawal to maintain the core critical was prohibited. Operations personnel were then directed to insert a manual scram signal.

Testing revealed the direct cause to be degraded signal cable shielding at under-vessel connectors in six of eight IRM channels allowing electrical noise to couple to the signal conductor. The noise was caused by a consistent low frequency signal on the preamplifier signal input and output cables and by degraded connectors. PM intervals were previously based on time rather than on duty cycle resulting in unidentified connector degradation.

Connectors were replaced and post maintenance testing confirmed noise had been reduced to acceptable levels. PM frequency changed based on duty cycle.

On April 16, 2011, during the Hatch Nuclear Plant outage, a local leak rate test (LLRT) was performed on torus purge supply primary containment isolation valve (PCIV) 2T48-F324 which is associated with primary containment penetration 2T23-X205. At that time, plant engineers and technicians were performing an LLRT for penetration 2T23-X205 when it was discovered that both PCIVs had failed their LLRTs for this penetration. This resulted in the penetration leakage exceeding the overall allowable leakage (La) required by the Technical Specifications (TS) for primary containment. This is considered a safety system functional failure of the primary containment function for this penetration.

The primary cause for the excessive leakage for valve 2T48-F324 was attributed to "over-travel" of the valve disc which reduced the seat contact with the disc. This was caused by inadequate procedural guidance for adjusting the valve travel. The excessive leakage for 2T48-F309 was attributed to the valve disc failing to "center" on the seat and wear on the actuator and its linkage. The necessary repairs and adjustments were made and the valves subsequently passed their respective LLRTs, which restored penetration 2T23-X205 to within its TS leakage limits. Additional procedural guidance is being provided to better control "travel" adjustments and in performance of valve inspections.