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Refueling Outage 22 commenced on January 14, 2017 at James A. FitzPatrick Nuclear Power Plant (JAF). With the plant in Mode 2 at 0613, the initial Drywell inspection identified a through wall leak on the 3/4 inch vent line off of the bonnet of the motor operated gate valve on the suction side of Reactor Water Recirculation Pump 'A'. This condition was determined to constitute Reactor Coolant Pressure Boundary (RCPB) leakage, which is prohibited by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.4. The average reactor coolant temperature decreased to less than 212 degrees F and the plant was in Mode 4 at 1530 of the same day, which is within the applicable TS LCO 3.4.4 required completion time.

The condition of a through wall leak on the RCPB is reportable pursuant to 10 CFR 50.73(a)(2)(ii), as a condition of the nuclear plant, including its principle safety barriers, being seriously degraded.

On June 24, 2016, at 1205, several 600V electrical busses lost power when James A. FitzPatrick Nuclear Power Plant was Reactor Water Recirculation (RWR). The 'A' RWR pump tripped immediately causing reactor power to reduce to approximately 50%. The remaining RBCLC pump was inadequate to maintain the MG Set fluid drive oil temperature for the 'B' RWR pump so Operators initiated a manual scram at 1236. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A).

The power loss also affected Reactor Building Ventilation (RBV). This system supports the requirement of Technical Specification Surveillance Requirement 3.6.4.1.1 for a differential pressure in Secondary Containment. At the loss of RBV, Secondary Containment automatically isolated and the Standby Gas Treatment system was manually initiated. However, during this short transition the differential pressure requirement was not met. This report is being submitted per 10 CFR 50.73(a)(2)(v)(C).

The apparent cause of the 71T-5 fault was inadequate preventative maintenance which allowed the transformer to remain in service beyond expected service life.

On January 23, 2016, James A. FitzPatrick Nuclear Power Plant (JAF) initiated a manual Scram in response to lowering screenwell water level due to frazil ice blockage, and subsequently closed the Main Steam Isolation Valves (MSIV). A post Scram review identified that MSIV 29A0V-8613 closed slowly. On January 26, 2016, testing per ST-1B identified that MSIV 29A0V-86C closed slowly. In both cases, the inboard MSIVs performed satisfactorily.

Troubleshooting identified that the problem originated in the solenoid valve cluster assemblies (SVCA) and they were replaced and tested successfully. A failure analysis was performed by Exelon PowerLabs on the SVCAs. On February 25, 2016, the Exelon PowerLabs analysis concluded that the DC pilot valves, 2950V-86B3 and 2950V-86C3, exhibited slow vent times. Additional corrective actions include changing the preventative maintenance frequency from 8 years to 6 years and initiating further investigation through the component's vendor.

Two MSIVs exceeded the closing time of Technical Specification Surveillance Requirement (SR) 3.6.1.3.6. This condition caused two independent channels of a system used to control the release of radioactive material to become inoperable; reportable per 10 CFR 50.73(a)(2)(vii).

On December 18, 2015, James A. FitzPatrick Nuclear Power Plant (JAF) was operating at 100 percent power when a 10 CFR 21.21(d)(3)(ii) Notification was received from Nutherm International. It identified a defect in Moore Industries temperature transmitters. Specifically, insulation was damaged in the T2 transformer during assembly which could result in premature failure.

These components were installed starting in June 2015 at 27TT-113A and 27TT-113B in the Containment Atmosphere Dilution (CAD) system. The defect caused failures in July and November which resulted in either the "A" or "B" CAD subsystem isolating. Corrective actions included replacing both temperature transmitters with ones that were confirmed to not contain this defect.

Even though these defective temperature transmitters function appropriately until they fail, this defect reduced the reliability of the CAD system to perform its function for its entire mission time. Therefore, this deficiency resulted in a loss of safety function to mitigate the consequences of an accident, reportable per 10 CFR 50.73(a)(2)(v)(D). Also, a single cause affected the safety function of independent CAD trains, reportable per 10 CFR 50.73(a)(2)(vii)(D); and, this condition existed longer then allowed by Technical Specifications 3.6.3.2, reportable per 10 CFR 50.73(a)(2)(i)(B).

At 2036 on December 1, 2015, James A. FitzPatrick Nuclear Power Plant (JAF) was operating at 100 percent power when differential pressure of the Secondary Containment exceeded a Technical Specification requirement. The requirement is for the Secondary Containment to have a >1= 0.25 inches of vacuum water gauge compared to the external environment.

This event was caused by hardened grease in the operating mechanism of the motor starter contactor 71MCC-141-062.

This led to a slow start of the above refuel floor Reactor Building exhaust fan 66FN-13B while the intake air supply fans were running. The exhaust fan started, without Plant Operator action, after approximately 60 seconds; however, Reactor Building differential pressure exceeded 0.25 inches of vacuum water gauge for approximately 80 seconds.

When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was Inoperable.

Restoration of the LCO was completed within the allowed action completion time. This report is being submitted per 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of safety function to control the release of radioactive material.

At 1408 on September 18, 2015, James A. FitzPatrick Nuclear Power Plant (JAF) was operating at 100 percent power when differential pressure of the Secondary Containment Reactor Building exceeded a Technical Specification requirement. The requirement is for the Secondary Containment to have a >1= 0.25 inch of vacuum water gauge compared to the external environment. During this event, Reactor Building differential pressure decreased below 0.25 inch of vacuum water gauge for approximately 3 minutes. This event was caused by the Refuel Floor Exhaust fan 'A' discharge damper 66A0D-106A going partially closed with concurrent failure of the associated fan discharge damper position indicating switch 66PNS-106A1. This resulted in obstructing exhaust flow from the Reactor Building. To correct the condition the standby exhaust fan was placed in service.

When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was Inoperable.

Restoration of the LCO was completed within the allowed action completion time. This report is being submitted per 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of safety function to control the release of radioactive material.

In February 2015 it was identified that the fuel support piece associated with fuel cell 38-39 was slightly elevated and offset.

As a result of the elevation and offset the four fuel assemblies in the affected fuel cell have reduced coolant flow; this has a direct effect on the Minimum Critical Power Ratio (MCPR). On April 30, 2015, an analysis of operating conditions by General Electric Hitachi (GEH) and accepted by the station confirmed that during periods of reduced reactor flow in October 2014 MCPR exceeded limits set by Technical Specification (TS). Since it was exceeded for longer than allowed by TS 3.2.2 this event is reportable as a condition prohibited by TS per 10 CFR 50.73(a)(2)(i)(B).

The fuel support piece became elevated during last refueling outage (September 2014). An interim corrective action to manually correct MCPR ensures the TS limit is not exceeded. The exceedance of the TS limit for MCPR was modified until the analysis was done since the core monitoring program assumes that the fuel support piece is in its normal seated configuration, with normal flow characteristics through the fuel assemblies.

Although the Limiting Condition of Operation was not met, the analysis performed by GEH confirmed compliance with Safety Limit MCPR under all evaluated transient scenarios such that at no time was the most limiting Safety Limit MCPR set by TS 2.1.1.2 challenged.

At 0630 on the morning of April 1, 2014, with James A. FitzPatrick Nuclear Power Plant (JAFNPP) operating at 100 percent power and the "A" Emergency Diesel Generator (EDG) subsystem inoperable for maintenance, a control room alarm annunciated indicating a problem with the "B" division safety pump room ventilation system. At 0645 a field operator identified the exhaust fan for the "B" division safety pump room ventilation system had tripped on thermal overload.

Ventilation loss when the pumps are in service would degrade the long term performance of residual heat removal service water (RHRSW) and emergency service water (ESW) systems; degradation of the "B" ESW pump would degrade the performance of the "B" EDG subsystem. The overload relay was reset at 0704 and the fan automatically started restoring ventilation; time out of service was thirty-four minutes. The ambient temperature limit in "B" division safety pump room was never challenged.

The equipment failure evaluation has determined preliminary cause of the tripping of the fan to be the weakening of the Bi- Metal trip element that is heated by the current that causes the trip. This component will be replaced and the unit tested to confirm cause.

On November 07, 2013, with the "A" Reactor Building Ventilation Radiation Monitor inoperable, the required action for Technical Specification (TS) 3.3.6.2 Condition A was not met within the required completion time of 24 hours. In addition, the required actions for TS 3.3.6.2 Condition C were also not met within the completion time of 1 hour after the action for Condition A was not completed. The failure to perform the TS required actions within the required completion times resulted in a condition prohibited by the TS which is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B).

The apparent cause of this event was operating procedures instructions were not adequately followed and understood. Additionally control room personnel failed to verify that the actual plant configuration matched the configuration needed to remain compliant with the TS. Immediate corrective actions included isolating the reactor building ventilation system to restore compliance with the TS. All involved control room personnel were removed from watch standing duties. Following remediation, some personnel were reinstated. Operations Management briefed on-coming watch crews on this event and the need to resolve discrepant items identified during turnover. Additionally, Shift Managers were briefed on the need to be more intrusive on TS related work preparation activities and execution.

Review of the as-found test results for eleven Safety / Relief Valve (S/RV) pilot assemblies removed and replaced during the September 2010 Refueling Outage, determined that five S/RVs were outside the allowable as-found tolerance of 1145 psig +/- 3% (+/- 34.3 psig) required by Technical Specification (TS) Surveillance Requirement (SR) 3.4.3.1. Also, two of the eleven S/RVs tested were found to have excessive seat leakage to the point where as-found testing could not be performed.

The effect of these S/RVs being out of tolerance was analyzed and the results of this analysis show that nuclear plant safety was not adversely affected due to the availability of the Electric Lift System.

Consequently, the safety significance of this event was minimal. The Root Cause for the failure of the S/RVs was determined to be corrosion bonding between the S/RV pilot disc and seat, a recognized industry generic problem with 2-stage Target Rock relief valves.

Safety valves in the Residual Heat Removal and Core Spray systems at the James A. FitzPatrick (JAF) Nuclear Power Plant failed to meet In-Service Testing (1ST) as-found acceptance criteria during surveillance testing conducted in February and March 2010. The most probable cause of the setpoint failure was the presence of internal binding and/or disc to seat corrosion bonding.

Component testing that falls outside the required action range for IST requires that the component be declared inoperable and the applicable Limiting Condition for Operation declared not met. 10 CFR 50.73(a)(2)(i)(B) allows licensees to consider the failure to have occurred at time of discovery unless there is firm evidence to indicate that the condition existed previously. Based on a review of test history, there is indication that this condition existed prior to discovery.

The event is reportable per 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications..." because during the time period described, the Residual Heat Removal and core spray subsystems were inoperable longer than allowed by TS. This event is also reportable per 10 CFR 50.73(a)(2)(vii), "Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains to become inoperable in a single system".

On September 16, 2012, James A. FitzPatrick Nuclear Power Plant (JAF) was reducing power for its scheduled Refuel Outage 20. During the power reduction, reactor pressure lowered below 800 psig prior to the full insertion of 52 control rods which were within the population for friction test control. These 52 control rods were conservatively declared Inoperable because friction testing was not performed within 14 days. This requirement is in accordance with the General Electric Part 21 notification, SC11-05 Revision 1, for Seismic Input in Channel-Control Blade Interference.

TS LCO 3.1.3 Condition E was entered for greater than 9 control rods declared Inoperable. This requires the plant to be placed in Mode 3 within 12 hours. Mode 3 was entered 2.5 hours later in accordance with the plan to shutdown for the Outage. The control rods remained fully functional and no control rod movement issues were experienced during the shutdown. An NRC notification was made per 10 CFR 50.72(b)(2)(i), initiation of any nuclear plant shutdown required by the plant's Technical Specification. This report is being made per 10 CFR 50.73(a)(2)(i)(A), completion of any nuclear plant shutdown required by the Technical Specifications. The SC11-05 compensatory actions will remain active. New fuel channels will be installed in future refueling outages to reduce the channel to control rod blade friction.