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On July 31, 2017 with Unit 1 at 100% Rated Thermal Power it was identified that Q1P17HV2229-Norm/Block Switch (Norm/Block Switch) was out of position (aligned to the "SI BLOCK" position). This is a key operated switch which controls the operation of valve Q1P17HV2229 (HV-2229), Component Cooling Water (CCW) Supply to Sample Coolers. HV-2229 functions to automatically isolate the non-seismic portion of the CCW system which includes the Reactor Coolant System (RCS) Sample Coolers. The Norm/Block switch allows a Safety Injection (SI) signal to be blocked to allow alignment of CCW to the RCS Sample Coolers for post-accident sampling. The Norm/Block Switch was returned to normal position on August 1, 2017 at 0415.

HV-2229 also receives a closed signal on low level CCW Surge Tank which will override the Norm/Block switch regardless of position. Thus, in the event of an SI concurrent with a CCW leak, the safety function of CCW would have still been met. Since HV-2229 was blocked from closing and would not have met Surveillance Requirement (SR) 3.7.7.2 from October 29, 2016 (entry to MODE 4) to August 1, 2017, the station unknowingly operated in a condition prohibited by Technical Specifications which is reportable under 50.73(aX2)(i)(B).

Corrective actions include procedure changes, communications and training to close knowledege gaps associated with system operation.

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3. UM NUMBER 1. FACILITY NAME

EVENT DESCRIPTION:

On July 31, 2017 with Unit 1 at 100% Rated Thermal Power en operator was preparing to lift a clearance order on the Component Cooling Water (CCW) System. The operator noticed that the Q1P17HV2229-Norm/Block Switch (Norm/Block Switch) was not in the correct position. The Norm/Block Switch was not part of the clearance order. The Norm/Block switch allows a Safety Injection (SI) signal to be blocked to allow alignment of CCW to the Reactor Coolant System (RCS) Sample Coolers for post-accident sampling. The operator notified shift supervision.

An investigation was conducted to identify when the Norm/Block Switch was last manipulated. Based on interviews, records, and plant conditions it was determined that the most probable date the Norm/Block Switch was operated occurred on October 1, 2016 during chemistry sampling following a reactor trip. Farley Unit 1 began a refueling outage following this reactor trip.

EVENT CAUSE ANALYSIS:

An analysis of the event identified organizational shortfalls in both procedure quality and human performance. The Chemistry operating procedure for obtaining the RCS sample did not provide a method of maintaining configuration control for the Norm/Block Switch. Procedural guidance on the Norm/Block Switch operation was only located in the precautions and limitations portion of the procedure and not in the instruction portion of the procedure. Additionally, communication between the Chemistry Technician and the Control Room Supervisor (key control) was not sufficient because the Chemistry Technician did not understand system status or realize that operation of the Norm/Block switch was not required for sampling since the SI signal had been reset. The Chemistry Technician operated the Norm/Block Switch based on knowledge of the switch function obtained from reading the precautions and limitations. The Chemistry Technician left the switch in the "SI-Block" position when returning the key to the control room. Upon return of the key to the control room no challenge was provided on the configuration of HV-2229 or the Norm/Block Switch.

REPORTABILITY AND SAFETY ASSESSMENT:

The auto closure of HV-2229 during an SI is surveilled per Surveillance Requirement (SR) 3.7.7.2. This SR requires that "each CCW automatic valve in the fiowpath that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal". The last scheduled performance of the surveillance testing of HV-2229 was performed April 4, 2015. Per SR 3.0.1, failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be a failure to meet the LCO. With the inability to automatically close HV-2229 on an SI signal due to the position of the Norm/Block Switch Farley Unit 1 unknowingly operated in a condition prohibited by Technical Specification upon entry into MODE 4 at 0632 on October 29, 2016. This condition is reportable under 50.73(8)(2)(0(B).

This event would not have prevented CCW from meeting its safety function and has very low safety significance because HV-2229 also receives a closure signal from a CCW Surge Tank Low Level signal. The Low Level Surge Tank signal overrides the Norm/Block Switch regardless of position. Therefore, this event is not reportable for loss of safety function per 50.73(a)(2)(v).

CORRECTIVE ACTIONS:

The cause of the event is that the manipulation of the Norm/Block Switch was not tracked using approved processes.

Barriers that broke down include; 1. Procedural guidance was not adequate for the task to maintain configuration.

2. Personnel knowledge level of the plant impact for performing the task did not meet standards.

3. Communications / Questioning attitude to obtain authorization were inadequate for the performance of the task.

4. The preparation and procedure use and adherence of the task was inadequate.

Farley has initiated corrective actions to address the organizational shortfalls and knowledge gaps associated with the event. This event was immediately communicated to the Department, Site and Fleet personnel. A training needs analysis is being conducted in Chemistry and Operations on the knowledge gaps and procedural guidance. The Site specific procedure for operation of the system has been revised to address configuration control of the Norm/Blodc switch.

PREVIOUS SIMILAR EVENTS:

None

OTHER SYSTEMS AFFECTED:

No systems other than those mentioned in this report were affected by this event.

(Joseph M. Farley Nudear Plant, Unit 1 05000- 348 2017 00

On 12/7/16 Units 1 and 2 were in Mode 1 at 100 percent rated thermal power. Engineering personnel determined that the Unit 1 and Unit 2 Service Water Intake Structure (SW IS) intake and exhaust ventilation hoods were not adequately protected from tornado generated missiles. Subsequent evaluations on 1/26/17 resulted in the determination that the emergency diesel generator (EDG) fuel oil storage tank (FOST) vents were also not adequately protected from tornado generated missiles. Operations subsequently declared the affected systems 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.

This comition is an original plant design legacy issue. Immediate compensatory measures included actions to take if needed as described in severe weather procedures, and monitoring of system parameters on the MCB in accordance with annunciator response procedures. A risk based evaluation will be performed or plant modifications will be undertaken to establish compliance with the site's tornado missile protection design basis.

On 11/1/2016 at 1743, Farley declared an Alert based on ammonia levels on the radiation side of the Auxiliary Building in the Recycle Evaporator room. The declaration was based on toxic gas emergency action level (EAL) HA3 which states, "Release of toxic, asphyxiant, or flammable gases within or contiguous to a VITAL AREA which jeopardizes operation of systems required to maintain safe operations or establish or maintain safe shutdown." The source was identified as a failed valve in Auxiliary Steam supply to the Recycle Evaporator system, which had previously been abandoned in place. The ammonia leak was subsequently isolated. Ammonia levels were reduced and the event was terminated at 2340 on 11/1/2016. The failure was determined to be due to ammonia- copper interaction in the brass valve material.

This report is being submitted under 10 CFR 50.73(a)(2)(x) for an event that hampered site personnel in performance of duties necessary for operation of the power plant due to a toxic gas release.

On November 8, 2016, Joseph M. Farley Nuclear Plant Unit 1 was reducing power to remove the main generator from service. The 1A steam generator feed pump did not respond to control steam generator levels as expected when the miniflow valve was opened per procedure. Steam generator levels lowered due to lower feed flow, and at 1331 the reactor was manually tripped from 32 percent power to prevent reaching the low steam generator level automatic reactor trip setpoint.

The motor driven auxiliary feed pumps also started automatically, as expected, with the manual reactor trip. The main steam isolation valves were closed to limit the unit cool down, decay heat removal was accomplished with the atmospheric relief valves, and the unit was maintained in mode 3. The controller failure was caused by the speed reference adjust and speed controller (C2) card being out of tolerance due to a failed A2 operational amplifier on the card, which was caused by infant mortality. The C2 card was replaced, and the new card was verified to be within the required tolerance. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) due to manual actuation of the reactor protection system and automatic actuation of the auxiliary feedwater system.

On 10/1/2016 at 0512 CDT with Unit 1 at 99 percent power the plant experienced a turbine trip and automatic reactor trip as a result of inadvertent closure of the 1 A Steam Generator Main Steam Isolation Valve (MSIV).

This caused a rapid pressure reduction in the remaining two Steam Generators' steam lines, resulting in a Safety Injection (SI). The 1 A MSIV closure was caused by failure of its test solenoid in conjunction with other air system leakage, which vented air pressure from the 1A MSIV actuator. An inadequate technical justification allowed the improper deactivation of the preventive maintenance (PM) task of the test solenoid valve in 2004. Decision making by control room personnel not to strictly adhere to an Annunciator Response Procedure was a contributing cause to the reactor trip being automatic versus manual, and led to the SI.

Following the reactor trip and SI the 1 A MSIV test solenoid was replaced and check valves on the 1 A MSIV steam line were tested and replaced. The technical justifications of a sample of previously extended or deleted PMs strategies will be reviewed and corrected. The PM for the Unit 1 solenoid will be reinstated.

Procedure use and adherence standards have been reinforced with Operations personnel, simulator just-in- time training was conducted for all crews, and further causal analysis is planned to investigate operations fundamental performance gaps. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) due to actuation of the reactor protection system, Emergency Core Cooling System (ECCS) injection into the Reactor Coolant System, and automatic actuation of the AFW system.

On November 12, 2015 at approximately 02:00 CDT, Farley Unit 2 was operating in Mode 5 at zero percent power in a planned maintenance outage when evidence from troubleshooting became available which led to the discovery that all of the Reactor Coolant System (RCS) Leakage Detection Instrumentation had been inoperable on August 7, 2015 for a period longer than allowed by Technical Specifications (TS).

Troubleshooting discovered that the A, B, and D channels of the Containment Cooling Level Monitoring System (CCLMS) had been inoperable since July 6, 2015. On August 7, 2015 the C CCLMS was tagged out at the same time as both the R11 and R12 Containment Radiation Detectors were taken out of service for a period longer than allowed by TS 3.4.15 Condition E and Limiting Condition for Operation (LCO) 3.0.3. This is a condition prohibited by Technical Specifications and is reportable in accordance with 10 CFR 50.73 (a)(2)(i)(B).

The cause of this event was an incorrect conclusion regarding the operating conditions of the four Containment Cooler Level Indicators. Corrective actions included the repair of the containment coolers' sensing lines for full restoration of the CCLMS. Repairs were also made to the components associated with the steam leak in containment N`RC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (It sots 19.."'.‘ I ) / LICENSEE EVENT REPORT (LER) APPROVED BY ()MD: NO. 3150.0104 EXPIRES: 10/31/2018 Reported lessons learned are incorporated into the licensing process and led back to industry.

Send comments regarding burden estimate to the FOIA, Privacy and Inform lion Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission. Washington DC 20555-0001, or by infernal e-mail to Infocollects.ResourceOnrc.gov, and to the Dal( Officer, Olice of Inlormalion and Regulatory Affairs, NEOB-10202. (3150-0104). Oka 01 Management and Budget.

Washington, DC 20503.11 a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required lo respond to, the information collection.

Joseph M. Farley Nuclear Plant, Unit 2 05000 - 364

NO

A. REQUIREMENT FOR REPORT

For a period of seven hours and 54 minutes on August 7, 2015 all required Reactor Coolant System Leakage Detection Instrumentation monitors required by Technical Specifications (TS) 3.4.15 were out of service and the required action per TS 3.4.15 Condition E and Limiting Condition for Operation (LCO) 3.0.3 to place the unit in Mode 3 within 7 hours was not met. This is a condition prohibited by TS and is reportable in accordance with 10 CFR 50.73 (a)(2)(i)(B).

B. UNIT STATUS AT TIME OF EVENT

Unit 2, Mode 1, 100 percent power

C. DESCRIPTION OF EVENT

On November 12, 2015 at approximately 02:00 CDT, Farley Unit 2 was operating in Mode 5 at zero percent power and was in a planned maintenance outage to investigate a leak ins'de of conta nment. A troubleshooting effort was initiated to investigate numerous maintenance issues that had been occurring with the A, B, and D Containment Cooler Level Indicators (EIIS Code LI). The troubleshooting work revealed that the A, B, and D Containment Cooler Level Indicators were inoperable and were considered to have been inoperable since July 6, 2015, when drainage into the containment sump had exceeded one gallon per minute. During this time frame the plant conducted extensive investigations into the source of the leakage, including multiple containment walkdowns and observations, maintenance troubleshooting activities, and detailed chemistry sample results of the containment sump, all of which led to the conclusion of a Service Water leak from the C Containment Cooler. The plant also prepared for a maintenance outage in the event that the leakage approached shutdown thresholds and implemented measures to protect the plant from the consequences of increased leakage.

During the November 2015 maintenance outage a steam leak was found in Containment and was verified as the cause of the Containment Cooling Level Monitoring System (CCLMS) alarms. The A, B, and D containment coolers' sensing lines were found to be clogged and therefore the level transmitters were unable to perform their function.

A subsequent review of operator logs showed that on August 7, 2015 the B and C Containment Cooler Level Indicators were declared inoperable to perform troubleshooting. On the same day, containment radiation monitors R11 and R12 were taken out of service for calibration surveillance. Since A and D CCLMS were discovered to have been inoperable, this created an unrealized entry into Tech Spec 3.4.15 Condition E (all detection inoperable) for a period of seven hours and 54 minutes, and the required action of immediate entry into Limiting Condition for Operation (LCO) 3.0.3 requiring the plant to be in MODE 3 in 7 hours was not met.

The CCLMS and the steam leak were repaired prior to exiting the maintenance outage.

Reported lessons learned are incorporated into the licensing process and led back to Industry Send comments regarcing burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mail to Infocollects.ResourceOnrc.gov, and to the Desk Officer, Office ot Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. II a means used to impose an irdormation collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond lo, the intonation collection.

Joseph M. Farley Nuclear Plant, Unit 2 05000 - 364

NO

D. CAUSE OF EVENT

The cause of the unrealized entry into Tech Spec 3.4.15 Condition E was an incorrect conclusion that the A, B, and D Containment Cooler Level Indicators, which had not been alarming, were operable, and that the C Containment Cooler Level Indicator was improperly alarming. This conclusion was reinforced by performance of multiple containment walkdowns and observations, maintenance troubleshooting activities, and detailed chemistry sample results of the containment sump that were strongly indicative of Service Water.

E. SAFETY ASSESSMENT

The leak in containment migrated to the containment sump which was monitored by radiation detectors.

The sump level was trended by a level monitoring indication. The site planned a maintenance outage to repair the leak. This condition had no significant effect on the health and safety of the public.

The loss of all CCLMS, along with a planned removal from service of R11 and R12 for calibration, represented an unplanned entry into Tech Spec 3.4.15 Condition E. The Condition requires an immediate entry into LCO 3.0.3 and entry into Mode 3 in 7 hours. The August 7, 2015 event lasted 7 hours and 54 minutes which exceeded the 7 hour time limit and therefore constitutes a condition that is reportable pursuant to 10CFR50.73 (a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications.

F. CORRECTIVE ACTION

During the November 2015 planned maintenance outage repairs were completed on the containment coolers' sensing lines for full restoration of the CCLMS. Repairs were also made to the components associated with the steam leak in containment.

G. ADDITIONAL INFORMATION

1) Failed Components: Level Indicator (LI) 2) Previous Similar Events: A search did not reveal any similar reported events for Plant Farley.

3) Energy Industry Identification System Code: Containment Leakage Control System (BD) 4) Other systems affected: There were no other systems, structures, or components that were affected by or contributed to the event.

5) Commitment Information: This report does not create any licensing commitments.

On 10/14/14 at 0341 CDT, Unit 2 reactor was manually tripped after a lightning strike In the High Voltage Switchyard (HVSY) led to a phase 3 to ground fault on a 500kV transmission line resulting in a B train Loss of Site Power (LOSP). The fault caused the 2B Startup Auxiliary Transformer (SAT) Instantaneous overcurrent relay to actuate and resulted in de- energizing the 2B SAT. A missing nut in the Power Circuit Breaker (PCB) protection circuitry caused a high resistance on one side of the current transformer circuit resulting in an imbalance in current flows and an actuation in the primary differential instantaneous overcurrent relaying. The B train LOSP in conjunction with the 2B Emergency Diesel Generator (EDG) being out of service for a planned maintenance outage caused a loss of Component Cooling Water (CCW) to the Reactor Coolant Pumps (RCP). The Unit 2 Abnormal Operating Procedures for loss of CCW and loss of A or B Train Electrical Power were entered and the reactor was manually tripped and the RCPs were secured. The reactor trip is reportable per 10 CFR 50.73(a)(2)(iv)(A) for manual actuation of the reactor protection system. Additionally, the reactor trip resulted in a valid actuation of the Auxiliary Feedwater system which is reportable per 10 CFR 50.73(a)(2)(iv)(A).

Corrective actions include: installed the missing PCB current transformer (CT) nut; satisfactorily tested primary and secondary protective relaying for 2B SAT; and strengthening of switchyard standards of the utility performing the maintenance. Extent of Condition walkdowns were performed for other circuits in the HVSY and repaired as necessary.

NBC FoRm 366 (03.2014)

On February 18, 2014 with FNP Units 1 and 2 operating at 100 percent thermal power, a review of industry operating experience determined that both units were in non-compliance with Technical Specification 3.4.3, Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits, during several previous refueling outages on each unit due to placing the RCS under vacuum conditions during RCS vacuum refill operations. From October 1995 through April 2012, 12 refueling outages have been conducted on Unit 1 and 11 on Unit 2 during which the RCS was placed under vacuum to perform vacuum refill operations. Technical Specification 3.4.3, applicable at all times, requires that RCS pressure, RCS temperature, and RCS heatup and cooldown rates be maintained within the limits specified in the Pressure Temperature Limit Report (PTLR). Although RCS temperature and heatup rates were maintained within limits, RCS pressure was lowered below zero pounds per square inch gage (psig), the lowest RCS pressure value identified on the curve.

The cause of not entering the required action for Technical Specification 3.4.3 was due to a failure to recognize that a negative RCS pressure is not allowed by Technical Specifications. An engineering review in support of the implementation of vacuum refill operations had previously determined that stress margins of the reactor pressure vessel and related components were not challenged. The station's PIT Limit curve is being revised to encompass RCS vacuum conditions.

On December 16, 2013, while Farley Units 1 and 2 were both operating at 100 percent power in Mode 1, a review of industry operating experience related to unfused direct current (DC) ammeter circuitry was concluded. This review determined that the condition of unfused DC ammeter circuitry was applicable to both Farley units and that a postulated fire could result in concurrent shorts to ground of a DC ammeter cable and a DC cable of opposite polarity. Due to a lack of overcurrent protection, the resultant excessive current flow in the ammeter cable could result in a secondary fire in another fire area. The secondary fire could adversely affect alternate safe shutdown capability contrary to 10 CFR 50, Appendix R, requirements.

This condition is reportable pursuant to 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety.

The cause of the condition was a latent design deficiency due to original plant design not providing DC ammeter circuits with overcurrent protection. Compensatory measures have been implemented for affected areas of the plant. Plant modifications are in development to provide these circuits with overcurrent protection.

On May 29, 2013, with Unit 2 operating in Mode 1 at 100% power, Engineering personnel performing a review of Unit 2 beginning-of-cycle power ascension data identified that 2C Steam Generator Steam Flow Transmitter FT-494 did not meet Technical Specification calibration accuracy requirements. Based on this information the steam flow instrument was declared inoperable and the required actions of the appropriate Technical Specification were performed.

However, since the data utilized in the engineering review was obtained on May 14, 2013, it is known that FT-494 has been inoperable since May 14, 2013. Consequently, the time limits of the applicable Technical Specification required action were not met. This represents a condition prohibited by Technical Specifications and is reportable under 10CFR50.73(a)(2)(i)(B). Steam flow transmitter FT-494 was re-calibrated and returned to service on June 1, 2013. This out-of- tolerance condition of FT-494 also occurred at the beginning of the previous fuel cycle. This supplemental report contains causal analysis and corrective action information that was not available for the original report.

Joseph M. Farley Nuclear Plant, Unit 2 05000 364

On February 15, 2012 at 0212, with Unit 1 operating at 100 percent power, 1B containment cooling fan (A Train) failed to start in slow speed during surveillance testing. Subsequent troubleshooting revealed that the failure of the fan to start was due to a malfunction of the 600 volt breaker that provides power to the fan motor slow speed winding. Based upon the investigation performed, it was determined the breaker had been rendered incapable of closing since January 18, 2012. This was the last time the breaker had been closed and the containment cooling fan had operated in slow speed satisfactorily. Although the 1B containment cooling fan was inoperable, the A Train of containment cooling was operable until the 1B containment cooling fan was selected for automatic start on January 23, 2012. Therefore, the A Train was inoperable for approximately 23 days until the failure was discovered on February 15, 2012. Since the seven day completion time of Technical Specification 3.6.6 Condition C was exceeded, this represents a condition prohibited by Technical Specifications and is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B).

The A Train operability was restored in approximately eight minutes when the 1A containment cooling fan was selected to automatically start on February 15, 2012 at 02:20. Investigation into the breaker failure determined that the cause was due to a failed charging motor cut-out switch.

NRC FORM 368 (10-2010) 2. DOCKET 6. LER NUMBER � I � 3. PAGE 1. FACILITY NAME 002 - 00 Joseph M. Farley Nuclear Plant. Unit 1 05000 348 2012

During the period between August 5 and August 9, 2009, the Unit 2 power supply to 1-2R 600 V load center (LC) did not meet the Unit 2 portion of Technical Specification (TS) 3.8.9 "Distribution Systems - Operating." This was discovered on August 9, 2009 when during routine surveillance for on-site AC distribution, the Unit 2 4160 V supply breaker DH08-2 to 1-2R 600 V LC was found in due to the 4160 V supply breaker DH08-2 from Unit 2 being open, and the required action statement was entered on August 9, 2009 at 02:02. Subsequently, the 4160 V supply breaker DH08-2 was closed on August 9, 2009 at 03:36. The 4160 V supply breaker DH08-2 being open resulted in a failure to meet TS 3.8.9 limiting condition for operation (LCO) for maintaining two trains of AC vital bus electrical power distribution subsystems operable. The 1-2R 600 V LC did not meet its surveillance requirement of correct breaker position and voltage for longer than the allowed by TS.

The 4160 V supply breaker DH08-2 was left open due to omission of relevant information in procedures and the interpretation of LCO 3.8.9 that existed at the time.