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On October 4, 2015 at 2317 hours, during the planned Unit 2 down power for entry into a refueling outage, the start-up feedwater (FW) pump failed to start. As a result, the motor driven FW pump (MDFWP) was used to supply FW to the steam generators for decay heat removal and cooldown.

On October 5, 2015 at 0038 hours, with the unit in Mode 3, MDFWP high journal bearing temperatures exceeded limits, and the pump was manually secured. Before the operating crew manually started the auxiliary feedwater (AF) system, the 2C steam generator low level reactor trip/AF actuation (Lo-2) setpoint was reached, resulting in an auto actuation of both trains of the AF system and an automatic reactor trip signal.

The cause of the failure of the startup FW pump to start was due to a pump starting interlock not making up. The corrective action planned is to create a post-maintenance test to perform an interlock check following maintenance of the startup FW pump.

The MDFWP elevated bearing temperatures were determined to be normal during low flow conditions experienced during the plant cooldown. Corrective actions completed included revising an operating procedure increasing allowable pump bearing temperatures.

On July 14, 2011, relic Asiatic clam shells were found in the essential service water suction piping between the cross-tie valves 2AF006A and 2AF017A, which provides a backup water source for the 2A Train auxiliary feedwater (AF) system.

At 1418, Technical Specifications Limiting Condition for Operation (LCO) 3.7.5, "Auxiliary Feedwater (AF) System," Condition A, "One AF train inoperable" was entered for the 2A AF train. Additional flushing of the suction piping was performed, and at 1151 on July 16, 2011, the 2A AF train was declared operable and the LCO exited.

The apparent causes for the event were: 1) Failure to perform a complete borescope inspection of the piping to verify no additional relic clam shells remained in the 2A AF suction piping from a May 9, 2011 event (reference LER 2011-001); and 2) Inadequate flushing techniques to remove shells from the 2A AF suction allowed for a significant amount of clam shells to remain in the system.

Corrective actions taken included flushing of the 2A AF suction piping between the 2AF006A and 2AF017A valves using improved flushing techniques, performing a complete borescope inspection of the 2A AF suction piping for remaining dam shells, and completion of the extent of condition flushing activities for the 1A, 1B and 2B AF trains.

There have been two similar Licensee Event Reports in the past three years: Unit 2 2008-001 and Unit 2 2011-001.

On December 27, 2008, at 14:18 hours, Braidwood Unit 2 Unit Aux Transformer 241-1 sudden pressure relay actuated causing a Unit 2 main generator trip, which resulted in a Unit 2 main turbine trip and subsequent Unit 2 reactor trip.

Concurrent with the reactor trip, the 2C heater drain pump (HD) tripped on phase "A" and "C" phase over current.

Operator response to the trip was proper and all safety related systems, structures and components operated normally during this event. The auxiliary feedwater system actuated, as expected, to maintain steam generator levels.

The investigation of this event determined the initiating event to the reactor trip was a phase-to-phase motor fault at the 2C HD pump motor terminal housing box, which caused a trip of the 2C HD pump on phase over current. Inspection of the motor lead box found the motor lead from one phase ("A" phase) in contact with the bus bar for another phase ("C" phase) due to excessive motor lead length. The root cause of the HD pump trip was determined to be that the procedure guidance for trimming the motor leads was deficient in that a lack of adequate information was provided for the desired motor lead length. The corrective action to prevent recurrence is to revise the existing procedure to provide clear direction on the desired length of power cables.

There were no actual safety consequences impacting plant or public safety as a result of this event.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to actuation of the reactor protection system (reactor trip) and the auxiliary feedwater system.

On October 2, 2016, during the liquid penetrant examination on the weld build up for control rod drive mechanism (CRDM) Penetration 69 during refueling outage Al R19, two rejectable rounded indications were documented. The first was a 7/32 inch rounded indication on the reactor head portion of the weld build up which was 4 inches from the transition of the head to penetration. The second was a 1/4 inch rounded indication located at the transition of the head to penetration.

The transition is the point where the vertical portion of the penetration meets the horizontal area of the reactor head. This LER is being submitted in follow-up to ENS 52275.

Based on industry experience, the cause of this event was determined to be mechanical discontinuities/minor subsurface voids opening up to the weld surface due to thermal and/or pressure stresses during plant operation.

The indications in penetration 69 were reduced to an acceptable dimension by manual buffing.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A), "any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded" since the as found indication did not meet the applicable acceptance criterion referenced in ASME Code Case N-729-1 to remain in-service without repair.

On March 4, 2016, during the NRC Component Design Basis Inspection, a concern was raised regarding why it was acceptable for the diesel driven auxiliary feedwater (AF) pump engine combustion air intake to be located in the turbine building, a non-safety related structure. On March 6, 2016, the additional evaluations that were completed determined that the existing configuration did not adequately support diesel engine operation with high energy line break (HELB) conditions in the turbine building, and at 2000 hours, Operations entered Technical Specification Limiting Condition for Operation 3.7.5, "Auxiliary Feedwater (AF) System," Condition A, "One AF train inoperable," for one train (B-train) of AF inoperable for both Units 1 and 2. The AF trains were declared operable following a corrective action to install a temporary configuration change to provide engine combustion air intake from the auxiliary building.

The cause of the event was insufficient validation of vendor analysis inputs in 1993 while reviewing the AF diesel engine's ability to function during a turbine building HELB event.

The corrective actions planned are to develop and install a permanent modification to re-route the AF diesel engine intakes for Unit 1 and 2.

examination was performed on the previously repaired control rod drive mechanism (CRDM) penetration 69. During the examination of the repair for CRDM penetration 69, one 3/8 inch rounded indication was documented exceeding the acceptance criterion (ASME Section III 1971 Edition through the Summer 1973 Addenda) of dimensions greater than 3/16 inch. This LER is being submitted in follow-up to ENS 50953 made on April 3, 2015.

Based on industry experience, the cause of this event was determined to be mechanical discontinuities/minor subsurface voids opening up to the weld surface due to thermal and/or pressure stresses during plant operation.

The indication in penetration 69 was reduced to acceptable size as approved by the NRC in Braidwood Relief Request I3R-09.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A), "any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded" since the as found indication did not meet the applicable acceptance criterion referenced in ASME Code Case N-729-1 to remain in-service without repair.

On February 19, 2014, it was determined that the Braidwood Generating Station has not complied with Technical Specifications (TS) 3.4.3, "RCS Pressure and Temperature (P/T) Limits," between March 2011 and October 2013, during start-up of the plant following plant refueling outages. Braidwood TS 3.4.3 Limiting Condition for Operation (LCO) states that "RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR." During previous Reactor Coolant System (RCS) vacuum fill operations at Braidwood Station Unit 1 and Unit 2, RCS pressure exceeded the Pressure and Temperature Limits Report (PTLR) P/T curve lower bound in that the PfT curve does not indicate a limit below 0 psig. This TS non-compliance is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition prohibited by the plants Technical Specifications".

The cause of operation outside of the P/T curve limits is the application of an inadequate operating procedure that allowed the P/T lower pressure bound to be exceeded during RCS fill operations.

RCS fill pressures below the PfT curve lower bound did not affect the integrity of the RCS system.

On August 1, 2013 at 1345, a through wall leak was observed on the Essential Service Water return line from the 1A Diesel Generator at a 90 degree horizontal elbow. Initial operability was determined to be supported based on the condition of the pipe and the assessment that the flaw was a localized pinhole.

It was later determined that based on the location of the flaw (on an elbow fitting), ASME Code Case 513-3 did not apply and Diesel Generator operability could not be supported. Due to the initial assessment of operability, Operations did not perform the Technical Specification (TS) 3.8.1 required 1 hour and 8 hour Diesel Generator availability TS Surveillance Requirements (SR) at the time of discovering the leak. The performance of the TS SRs is to verify the correct breaker alignment and indicated power availability for each required qualified circuit.

The cause of this event was determined to be the complexity of applying the procedure OP-AA-108-115, "Operability Determinations" section on ASME flaw evaluation/Class 1, 2 and 3 leakage.

Corrective actions to prevent recurrence include revising the Operability Determination procedure, OP-AA-108-115, to provide clearer guidance when applying the requirements of ASME Code Case 513-3,

On February 25, 2011, dried boric acid was identified on a 1A safety injection (SI) pump discharge line. At 1830 hours, upon receipt of the non-destructive examination results indicating a potential pressure boundary leak, the lA SI train was declared inoperable, and Technical Specification Limiting Condition for Operation (LCO) 3.5.2, "Emergency Core Cooling Systems - Operating," Condition A was entered for one train inoperable. Following pipe replacement, on March 3, 2011, at 2028 hours, the system was returned to service and LCO 3.5.2 Condition A was exited.

The apparent cause of the through wall crack is outside diameter (transgranular) stress corrosion cracking that initiated from the external surface of the pipe caused by chloride exposure.

Corrective actions included replacing the portion of the SI line containing the flaw with new stainless steel pipe, and performing extent of condition walkdowns of portions of the SI discharge piping.

There were no actual safety consequences impacting plant or public safety as a result of the event.

This event is reportable under 10 CFR 50.73(a)(2)(i)(B), any operation or condition which is prohibited by the plant's Technical Specifications.