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 Report dateSiteEvent description
05000456/LER-2016-00323 November 2016Braidwood

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.

05000456/LER-2016-00219 July 2016Braidwood

On May 25, 2016, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornadoes, Braidwood identified non-conforming conditions in the plant design such that specific TS equipment on both units was considered to not be adequately protected from tornado missiles.

On May 25, 2016 at 1415 Operations declared the affected equipment inoperable, implemented Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance" and the required compensatory measures, and then declared the affected equipment operable but non-conforming.

The cause of this issue was a lack of clarity and changing requirements during the original licensing of the plants which led to inadequate understanding of the original NRC regulatory guidance.

The corrective actions planned are to complete the EGM 60-day comprehensive compensatory measures to demonstrate a discernable change from its pre-discovery actions, to modify the refueling water storage tank hatches to eliminate the tornado missile vulnerability, and to obtain and implement a license amendment for an analytical solution dispositioning tornado generated missile nonconforming conditions.

05000456/LER-2016-00128 April 2016Braidwood

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.

05000457/LER-2015-0024 December 2015Braidwood

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.

05000456/LER-2015-00319 October 2015Braidwood

On August 20, 2015, a design deficiency associated with the pressurizer power operated relief valve (PORV) block valve control circuitry was confirmed, in which a design basis fire in the main control room or cable spreading rooms could prevent the credited fire safe shutdown action (i.e., locally close pressurizer PORV block valve) mitigating a spurious pressurizer PORV opening.

On September 2, 2015, during an extent of condition review, an additional design deficiency was confirmed in which credited fire safe shutdown action (i.e., removing control power fuses) mitigating a spurious opening of the pressurizer PORVs during a design basis fire, does not adequately mitigate a design basis fire induced hot short.

The causes of these design deficiencies are legacy design errors made during original construction.

Corrective actions include plant configuration changes correcting the specific design deficiencies.

This condition is being reported in accordance with 10CFR50.73(a)(2)(ii)(B), any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

05000457/LER-2015-00324 August 2015Braidwood

storage tank (DOST) fuel oil level increased. This DOST level change was caused by leak by of the 2A DOST fill valve (2D0001A). Engineering evaluated the 2B DG fuel oil volume with consideration for the 2D0001A leak by and determined that the 2B EDG did not have sufficient fuel oil to satisfy its 7 day mission time specified in Technical Specification (TS) 3.8.3 "Diesel Fuel Oil".

This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B), for "any operation or condition which was prohibited by the plant's Technical Specifications" since the 2B DG was not declared inoperable and some associated Limiting Condition For Operation Action Requirements (LOCAR) were not completed.

The cause of the event was a degraded DOST fill valve that allowed fuel oil to transfer from the 2B DOST to the 2A DOST during 2B DG operation.

Corrective actions include repairing of 2D0001A and implementing a DOST fill valve leakage monitoring program with associated acceptance criteria, or implementing a plant configuration change to install an additional valve in the flow path.

05000456/LER-2015-0022 June 2015Braidwood

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.

05000457/LER-2015-00127 March 2015Braidwood

On January 26, 2015 while performing a quarterly surveillance stroke test of the 2C Steam Generator (SG) Power Operated Relief Valve (PORV), 2MS018C, the associated SG PORV manual isolation valve, 2MS019C, failed to close.

The last successful quarterly stroke of 2MS019C was performed on October 27, 2014.

A failure analysis performed concluded that moisture had been present in the valve's gear box for an extended period of time, causing corrosion and pitting in the 2MS019C actuator. The degradation removed enough material to cause the remaining intact cross-sections to fail during the attempt to close the valve on January 26, 2015.

There is the potential that a condition prohibited by technical specifications (TS) existed before discovery, for a time longer than permitted by TS, based on the surveillance frequency. This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications." Corrective actions included repairing the valve, performing a failure analysis of the actuator gears from 2MS019C, and performing an extent of condition review on similar valves.

05000456/LER-2015-00112 March 2015Braidwood

On January 13, 2015, Braidwood Station was reviewing operating experience (OE) related to Main Steam Isolation Valve (MSIV) operability and changed its previous position regarding MSIV operability. Previous MSIV operability practice considered a MSIV operable as long as one of two redundant actuator trains was operable. Revised MSIV operability practice requires both redundant MSIV actuation trains operable. Declaring a MSIV inoperable and entering MSIV TS LCO when single MSIV actuation train is inoperable is consistent with an October 19, 2006 NRC staff interpretation for different plant under a Task Interface Agreement (TIA). An extent of condition review identified two occurrences in the previous three years when a single MSIV actuator train was inoperable and not restored within the required TS 3.7.2 eight hour completion time. In these instances, TS 3.7.2 requirements were inadequately applied under the new interpretation and should have been imposed. This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications".

The cause of the event was the legacy MSIV TS and TS Bases did not explicitly address MSIV actuator trains. Corrective actions included communicating to operating crews, procedure revisions and implementing new MSIV TS providing actions for inoperable MSIV actuator trains.

05000456/LER-2014-00117 April 2014Braidwood

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.

05000456/LER-2013-00213 November 2013Braidwood

On September 14, 2013 at 0100 hours, indications exceeding ASME Section III acceptance criteria were discovered in the embedded flaw repair weld of Control Rod Drive Mechanism (CREW) Penetration 69. A total of 22 recordable indications located within the embedded flaw repair weld were documented, with 13 rounded indications exceeding the acceptance criteria (ASME Section III 1971 Edition through the Summer 1973 Addenda) of dimensions greater than 3/16. The indications in Penetration 69 were removed/reduced to acceptable size or reduced and weld repaired to restore the embedded flaw repair as approved by the NRC in Braidwood Relief Request 13R-09. This LER is being submitted in follow-up to ENS 49343 that was made at 0447 hours on September 14, 2013.

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.

Corrective actions to prevent recurrence include consideration of a long term Alloy 600 mitigation plan and other potential alternatives for the remaining acceptable but unrepaired penetrations on the Braidwood Units 1 and 2 RPV closure heads. This corrective action plan is under development and will also evaluate peening of the remaining acceptable but unrepaired CRDM penetrations.

05000456/LER-2013-00130 September 2013Braidwood

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,

05000457/LER-2011-00212 September 2011Braidwood

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.

05000457/LER-2011-00119 July 2011Braidwood

On May 9, 2011, Asiatic clam shells were found in the essential service water (SX) suction piping between the cross-tie valves 2AF006A and 2AF017A (SX, through these valves, provides a backup water source for the Train 2A auxiliary feedwater (AF) system in the event the normal water source for the AF system, becomes unavailable).

On May 20, 2011, the evaluation of past operability concluded Train 2A of AF was not operable with this quantity of shells. The clam shells which were identified in the suction piping between the 2AF006A and 2AF017A valves had the potential to be transported through the 2A AF system and block flow through the AF flow control valves (2AF005A-D).

The root causes for the event were determined to be: 1) Historical inadequate chemical feed biocide treatment of SX prevented the termination of Asiatic clam larvae, allowing Asiatic clams to grow in the low flow section of the SX header to 2A AF suction; and 2) Historical ineffective Problem Identification and Resolution associated with discoveries of Asiatic clam shells in the 2A AF suction allowed the condition to remain unanalyzed and uncorrected.

Corrective actions include flushing the line between the 2AF006A and 2AF017A valves, revising the service water heat exchanger inspection guide to incorporate additional guidance on actions to be taken upon the discovery of macrofouling and biological fouling of NRC Generic Letter 89-13 systems, and revising the AF 1/2AF006A/B and 1/2AF017A/B valve stroke surveillance to clarify where and how to document the discovery of debris in the system.

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

05000456/LER-2011-0023 May 2011Braidwood

On October 11, 2010, during execution of an undervoltage relay surveillance on Train A of the 4.16 kV Engineered Safety Feature bus 141, the as-found trip values of the A-to-B phase degraded voltage relay were found out of tolerance. The relay was replaced.

An evaluation identified that the undervoltage relay had a manufacturing defect. A polarity sensitive capacitor was incorrectly installed with the polarity reversed.

On March 4, 2011, a review of this evaluation was initiated to identify when the instrument out of tolerance occurred.

Based on a review of this event against known conditions and trends, there was insufficient evidence to predict when the relay failure would have occurred, and no past surveillance history would indicate degraded conditions prior to the out of tolerance condition. Therefore the relay out of tolerance issue may have existed for a longer period of time than is allowed by the plant's Technical Specifications.

Corrective actions included replacement of the degraded voltage relay.

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.

05000456/LER-2011-00126 April 2011Braidwood

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.

05000457/LER-2008-00225 February 2009Braidwood

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.