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On March 14, 2015 at 1207 Eastern Daylight Time (EDT), Unit 2 was operating in Mode 3 following an unplanned reactor trip and safety injection, a valid Auxiliary Feedwater (AFW) actuation signal was received on B-train Auxiliary Feedwater system during post-trip recovery and unit stabilization. Both Train A and Train B motor driven AFW pumps were in service at the time. Upon receipt of the AFW actuation signal the B-train discharge valves stroked to the full open position. Operators immediately restored AFW discharge valves to their previous positions without any adverse impacts on the unit.

The plant remained in Mode 3 and decay heat continued to be removed via the Atmospheric Relief Valves.

Unit 1 was unaffected. This event had no adverse effect on the health and safety of the public, and is of very low safety significance. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) due to actuation of a system listed in 10CFR 50.739a)(2)(iv)(B).

On July 27, 2014 at approximately 1409 Eastern Daylight Time, Vogtle Unit 1 was operating in Mode 1 at 100 percent power. While surveillance testing was being performed on the Main Feed Pump A (MFP A) a control oil leak developed in the lockout solenoid valve. The reduced control oil pressure resulted in a reduction of feed pump speed. Control Room Operators observed lowering feedwater flow to the Steam Generators and manually actuated the Reactor Protection System (RPS) system which resulted in a turbine-generator trip. All rods fully inserted into the core, the Main Feedwater Isolation system and the Auxiliary Feedwater system automatically actuated as expected. The unit was stabilized in Mode 3 and decay heat was discharged to the condenser. The cause of the event was a failure of the trip lockout solenoid valve which led to reduced pressure on the trip relay for MFP-A which reduced the speed of the pump and consequently the total feed flow to the Steam Generators.

The safety significance of the event is very low. Unit 2 was not affected and there were no adverse effects on the health and safety of the public.

APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 Estimated burden per response to corn* Mth this mandatory collection request 80 hours.

Reported lessons teamed are incorporated into the Ficensitv process and fed back to industry.

Send comments regardng burden estimate to the FOIA. Rimy and Information Collections Branch (T-5 F53). U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NE013-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not dsplay a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

On November 27, 2013, an internal wiring discrepancy was discovered on Class 1 E battery charger 1AD1CB (EJ) following the trip of the battery charger AC input breaker.

The wiring discrepancy prevented the battery charger from performing all required functions of LCO 3.8.4, DC Sources - Operating. One of two redundant battery chargers and one battery per train must be operable to meet the requirements of LCO 3.8.4. Subsequent review of battery charger maintenance activities determined that on September 30, 2013 the remaining redundant operable battery charger was removed from service for maintenance activities for approximately 14.5 hours.

LCO 3.8.4, DC Sources - Operating requires restoration of the inoperable DC source within 2 hours or entry into Mode 3 within the next 6 hours, and entry into Mode 5 within the following 30 hours. Although the degraded charger was able to maintain battery terminal voltage within limits under minimal loading condition, the DC source was inoperable for a time greater than allowed by Technical Specification.

The safety significance of this event is very low. Unit 2 was not affected and there were no adverse effects to the health and safety of the public.

NRC FORM 365(104010) NRC RM 365I1 004010) FO LICENSEE EVENT REPORT (LEI) U.S. NUCLEAR REGULATORY COMMISSION

On February 15, 2012, with the unit at 100 percent power, it was determined that opening the boundary valve between the safety related and seismically qualified Refueling Water Storage Tank (RWST) and the non safety related and non seismically qualified Spent Fuel Pool Purification (SFPP) system in Modes 1-4, renders the RWST inoperable. Plant procedures had been revised In 2009 to allow opening this boundary valve in Modes 1-4 under administrative controls. The 10 CFR 50.59 safety evaluation that had been performed to support the procedure change had concluded that the administrative controls would allow the RWST to remain operable. However, in consideration of the new interpretation provided in NRC Information Notice 2012-01, it was judged that the RWST would be considered to be inoperable regardless of the administrative controls established when the RWST was aligned to non-seismic piping in Modes 1 - 4. Since the boundary valve had been opened in Mode 1 under administrative controls and the one hour completion time of Technical Specification 3.5.4 Condition D was not entered, under this recent interpretation, this represented a condition prohibited by Technical Specifications and is reportable pursuant to 10 CFR 50.73(a)(2)(I)(B). This event had no significant safety consequence since a seismic event had not occurred while the SFPP system was in service on the RWST.

NRC FORM 386 (10.2010)

On August 23, 2010 an Operations Superintendent was reviewing past performances of the reactor trip breaker surveillance test in preparation for an upcoming surveillance test on a reactor trip breaker.

During this review it was determined that on January 28, 2010 the 2B Solid State Protection System (SSPS) Mode Selector switch was placed in Test to perform a surveillance test on the 2B reactor trip breaker. Placing the 2B SSPS Mode Selector switch in Test defeats the Engineered Safety Features (ESF) automatic actuation signal to the 2B train components. At the time 2B SSPS Mode Selector was placed in test, the 2A high head safety injection (HHSI) pump was tagged out of service for planned maintenance. In this configuration, both trains of HHSI were rendered inoperable and outside the Conditions stated in Technical Specification 3.5.2, since neither train would have automatically started on an ESF actuation signal. Although not recognized at the time of the event, this condition required entry into Technical Specification LCO 3.0.3.

The cause of this event was due to an inadequate Loss of Safety Function (LOSF) determination, as described in Technical Specification 5.5.15, being performed and the work planning process not adequately identifying potential Technical Specification implications for the scheduled work.

hNRC FORM 366 (9-2007) PRINTED ON RECYCLED PAPER 5

The week of June 6, 2005, a Vogtle Equipment Qualification Program self-assessment was performed that identified a potential problem with the Rosemount model 1153 Series B and 1154 transmitters if the neck seal is broken. This seal protects the transmitter electronics from moisture intrusion to ensure the safe operation of the transmitter during accident conditions. This self-assessment resulted in the development of an inspection plan to determine if this condition existed at Vogtle. The inspection plan was initiated in June 2005 and completed in August 2006. The following three Technical Specification (TS) instruments were determined to have been inoperable as a result of a broken neck seal: 1FT-5152, Steam Generator (SG) 1 Auxiliary Feedwater Flow, 2PT-0455, Pressurizer Channel 1 Pressure, and 1PT-0456, Pressurizer Channel 2 Pressure.

Per the vendor manual, the connection between the electronic housing and the sensor module is hermetically sealed with a baked-on environmentally qualified neck seal. The cause of the event was a result of inadequate installation and calibration procedure guidance in that a caution statement from the vendor manual not to break the neck seal between the sensor module and the electronics housing was not included. All three transmitters were replaced, and appropriate maintenance procedures have been updated to reflect the vendor manual caution to not rotate the head.

A.,,, rAllial IA& illitilllidn �NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001) Vogtle Electric Generating Plant - Unit 1 05000-424 2006 --� 002 --� 00 20F5

On February 1, 2006, control room operators received indication of an increase in radioactivity in the containment atmosphere. On February 3, 2006, a robotic camera observed leakage inside the bioshield wall in the area of reactor coolant system (RCS) loop 1. Unit 2 was placed in Mode 3 (Hot Standby) at 1806 EST, on February 3, 2006, to allow further investigation of specific leakage locations. At 2124 EST, this investigation found RCS pressure boundary leakage at two welded connections on a 'A" bypass line around the Residual Heat Removal (RHR) loop suction valve, 2HV 8701B, and shutdown to Mode 5 (Cold Shutdown) was initiated. On February 5, 2006, at 0035 EST, Unit 2 entered Mode 5 to comply with Technical Specification 3.4.13.a. due to the RCS pressure boundary leakage.

Although the cause of the weld failures is undetermined at this time, the root cause investigation is still underway. Issues being addressed include high vibration of the bypass line, support design, excessive weld stress, and weld quality. The theory presented by the root cause team is the cause of the cracks was from high cycle fatigue. Corrective actions included replacement of the Loop 1 bypass line, inspection of supports and snubbers for both Unit 2 bypass lines, inspection of welds on the Loop 4 bypass line, and installation of monitoring instrumentation. Following additional monitoring, the cause determination will be completed, follow-up corrective actions implemented and a revised LER will be submitted by July 11, 2006.