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 Report dateSiteEvent description
05000333/LER-2017-00113 March 2017FitzPatrick

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.

05000333/LER-2016-00517 November 2016FitzPatrick

On September 19, 2016, the Control Room Ventilation exhaust fan 70FN-4A did not start when it was being placed into service. The fan outlet isolation damper actuator 70MOD-108A(OP) failed to give the fully-open permissive signal to start the fan. Gentle pressure on the actuator linkage allowed the fan to start. Prior to this, on August 16, 2016, during post-maintenance testing, 70FN-4A did not start. Troubleshooting adjusted the linkage and the fan started as appropriate. However, the intermittent fan start issue was caused by the degraded damper actuator 70MOD-108A(OP). Corrective action replaced 70MOD-108A(OP).

This event is reportable per 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.

05000333/LER-2016-00423 August 2016FitzPatrick

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.

05000333/LER-2016-0033 August 2016FitzPatrick

On the morning of June 7, 2016, while operating at 100% power, workers opened doors concurrently when entering a secondary containment access airlock. The individuals involved each closed their respective doors upon encountering this unexpected condition; however, the result was a brief inoperability of secondary containment.

This resulted in an 8 hour reportable event. The Resident Inspector was notified, and an Event Notification was made pursuant to 10 CFR 50.72(b)(3)(v)(C) due to a condition at the time of discovery that prevented the fulfillment of the Secondary Containment safety function (Reference ENS 51985). Following the event, the doors functioned properly, and no deficiencies were noted with either door.

There were no radiological releases associated with this event.

05000333/LER-2016-00225 April 2016FitzPatrick

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).

05000333/LER-2016-00123 March 2016FitzPatrick

On January 23, 2016, James A. FitzPatrick Nuclear Power Plant (JAF) was ascending in power when screenwell water level started to lower. At 89 percent power, at 22:23, Operators began taking compensatory measures to reduce power and mitigate water level lowering. At 22:40, a manual scram was initiated.

The scram was complicated by a residual transfer that resulted in non-vital equipment trips. This event resulted in the manual actuation of the Reactor Protection System, High Pressure Coolant Injection, Reactor Core Isolation Cooling, Main Steam Isolation Valves and automatic actuation of Emergency Diesel Generators, Emergency Service Water, and containment isolations in multiple systems, reportable per 10 CFR 50.73(a)(2)(iv)(A).

The lowering screenwell water level was caused by frazil ice blockage at the intake structure. The frazil ice stopped affecting screenwell water level after the manual scram. Corrective actions include strengthening mitigating actions in response to frazil ice.

The residual transfer was caused by lubrication hardening in the lower control valve assembly of the 71PCB-10042 breaker. Corrective actions included replacing or reworking the lower control valve assembly.

05000333/LER-2015-00816 February 2016FitzPatrick

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).

05000333/LER-2015-0064 February 2016FitzPatrick

On September 22, 2015 at 17:03, with James A. FitzPatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated a spike in Secondary Containment (SC) differential pressure (d/P) during performance of a surveillance test associated with automatic isolation of SC and initiation of the Standby Gas Treatment System. Per the plant data systems SC d/P exceeded the Technical Specification (TS) allowed value, and then immediately trended negative following auto-start of one of the trains of Standby Gas Treatment.

The time period that SC d/P was greater than the TS allowed value is reportable pursuant to 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.73(a)(2)(v)(C), as an event or condition that could have prevented fulfillment of a safety function. SC was operable following reestablishment of greater than or equal to 0.25 inches of water vacuum, and remains operable.

SC d/P excursions during transition from normal to isolation mode of the Reactor Building Ventilation (RBV) System are an expected condition, and attributable to the design of the non-safety related RBV System. The cause of the SC d/P exceeding the TS allowed value has been determined not to be associated with a component failure or equipment malfunction. Similar reportable events were identified during preparation of this report. A comprehensive listing of these occurrences is included in the report.

05000333/LER-2015-0071 February 2016FitzPatrick

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.

05000333/LER-2014-00228 December 2015FitzPatrick

At 1655 and 1708 on October 28, 2014, 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. The Reactor Building differential pressure decreased below 0.25 inch of vacuum water gauge twice: at 1655 when isolating Reactor Building Ventilation caused by the isolation valve closing sequence; and, at 1708 while restoring Reactor Building 'A' Ventilation 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. Each condition existed for approximately a minute.

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.

05000333/LER-2015-00517 November 2015FitzPatrick

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.

05000333/LER-2015-00318 September 2015FitzPatrick

On the morning of July 20, 2015 at 0740 EDT, with the James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, the differential pressure (D/P) of the Reactor Building (relative to atmosphere) decreased, resulting in a Technical Specification (TS) surveillance requirement for the Secondary Containment vacuum not being met (>1= 0.25 inches of vacuum water gauge). The decrease in D/P was caused by the opening of the Reactor Building roof envelope during roof replacement activities. The condition of D/P being less negative than 0.25 inches of vacuum water gauge existed for approximately 92 minutes.

When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the TS Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was Inoperable. Restoration of the LCO was completed after secondary containment was declared Operable / Degraded. 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.

05000333/LER-2015-00129 June 2015FitzPatrick

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.

05000333/LER-2012-00126 June 2014FitzPatrick

On 1/26/12, a first time low voltage pickup test was performed for 71MCC-163-OE5 (East Crescent Unit Cooler 66UC-22H Fan Motor). The contactor picked up at 102 VAC versus the 90 VAC Level 2(1) Acceptance Criteria. A calculation was performed to determine Level 1(1) Acceptance Criteria for this application. The Level 1 Acceptance Criterion was established as 97 VAC. The Unit Cooler was therefore non-functional, and is assumed to have been non-functional for a period of three years prior to the time of discovery. 66UC-22H is one of five Unit Coolers in the East Crescent Area Ventilation Subsystem.

A review of historical plant data identified five occurrences where the non-functional 66UC-22H resulted in the East Crescent Area Ventilation Subsystem being incapable of performing its specified function, and therefore non-functional. The East Crescent Area Ventilation Subsystem is required to support operability of the ECCS equipment located in the East Crescent. This ECCS equipment was therefore inoperable during the five time periods that were identified. This condition is prohibited by plant Technical Specifications, and resulted in the loss of safety function for the single train HPCI System.

Revision 1 is submitted as clarification, and to correct inaccuracies in the original report relative to how the reporting requirements of 10 CFR 50.73(a)(2)(i)(B) are met by this condition.

05000333/LER-2014-0012 June 2014FitzPatrick

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.

05000333/LER-2013-00618 February 2014FitzPatrick

On the morning of December 17, 2013, at James A. FitzPatrick (JAF) Nuclear Power Plant, with the reactor at 100 percent power, performance of ISP-75 commenced at 0810 EST. ISP-75 functionally tests and calibrates the Condensate Storage Tank (CST) Low Water Level instrument channels of the High Pressure Coolant Injection (HPCI) System. During performance of this planned surveillance, the first level switch associated with the "B" CST that was tested (23LS-74B) failed to actuate. At 0938 EST, the Chief Instrumentation and Controls technician performing the test reported the failure of 23LS-74B to the Control Room Supervisor, and was directed to continue testing per ISP-75. Subsequent testing of the redundant "B" CST level switch (23LS-75B) revealed that the level switch actuated at a simulated CST level of 58.5 inches.

The combination of these two deficiencies would have prevented the HPCI automatic suction swap-over function until the "B" CST level dropped to 58.5 inches. This is less than the analyzed Technical Specification (TS) required value for low CST water level of 59.5 inches; per TS 3.3.5.1 Condition D, this would have resulted in the HPCI system being declared inoperable. Therefore, this event is reportable pursuant to 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of a safety function.

05000333/LER-2013-0056 January 2014FitzPatrick

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.

05000333/LER-2011-00115 November 2011FitzPatrick

On January 7, 2011, with the plant operating in Mode 1 at 100% power, 13MOV-131, Reactor Core Isolation Cooling (RCIC) Steam Admission Isolation Valve failed to stroke full open during surveillance test ST-24J. Troubleshooting determined the most probable cause to be loose connections in the motor control circuit, 71BMCC-3-0B1(MC).

Preventive Maintenance (PM) had been performed on the motor control circuit on September 23, 2010, during Refueling Outage 19. Because the identified condition could have resulted in a failure of the RCIC system to operate properly, if needed, it is considered that the RCIC System was Inoperable from the time that RCIC was required to be Operable on October 16, 2010, until the completion of the Post Maintenance Testing on January 8, 2011. Since Limiting Condition for Operation (LCO) 3.5.3 requires RCIC to be Operable in Mode 1 and in Modes 2 and 3 with steam dome pressure greater that 150 psig, this period of Inoperability exceeded the Technical Specification allowed out of service time.

05000333/LER-2011-0038 August 2011FitzPatrick

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.

05000333/LER-2011-00210 March 2011FitzPatrick

On January 11, 2011, with the "A" Refueling Floor Exhaust 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 is that 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 and briefing the operating crews on the error prior to assuming future watch standing duties.

The briefings were conducted by operations management who also reinforced expectations regarding shirt turnover and T5 compliance. Additionally, active LCOs are now visibly posted in the control room. Planned corrective actions include providing additional training and simulator scenarios on configuration control.

05000333/LER-2010-00120 May 2010FitzPatrick

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".

05000333/LER-2009-00522 June 2009FitzPatrick

Review of the as-found test results for 11 Safety / Relief Valve (S/RV) (SB) pilot assemblies removed and replaced in September 2008, determined that 5 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.

The effect of these S/RVs being out of tolerance was analyzed and the results of this analysis show that Reactor Pressure Vessel (RPV) overpressure protection and nuclear plant safety were not adversely affected. Consequently, the safety significance of this event was minimal. The most probable cause for the failure of four of the S/RVs was determined to be corrosion bonding between the S/RV pilot disc and seat, a recognized industry generic problem with two-stage Target Rock relief valves. The fifth failure was determined to be due to significant pilot valve seat leakage which would have required additional steam pressure to overcome the leakage and lift this S/RV.

05000333/LER-2012-0041 January 1111 JLFitzPatrick

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.