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On June 30, 2016 at 1430 (EDT), with the reactor at 100 (percent) and the mode switch in RUN, Pilgrim Station determined both Primary Containment Isolation Valves (PCIVs) CV-5065-91 and CV-5065-92 for Drywell Penetration X-32A were inoperable due to the potential failure of relays relied on to perform the primary containment isolation function. The valves have been closed and deactivated in the isolated condition in accordance with Technical Specification Limiting Condition For Operation Action Statement 3.7.A.2.b. Preparations are in progress to replace the relays to restore the valves to operable status. This 8-hour notification is being made in accordance with 10 CFR 50.72 (b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D). The licensee has notified the NRC Senior Resident Inspector. The licensee will notify the Commonwealth of Massachusetts.

• * * RETRACTION FROM KEN GRACIA TO DONALD NORWOOD AT 1720 EDT ON 8/29/2016 * * *

This notification is being made to retract event notification EN 52056 made by Pilgrim Nuclear Power Station on June 30, 2016, that reported the potential failure of relays that could have prevented the fulfillment of the safety function of primary containment isolation valves (PCIVs) needed to control the release of radioactive material and mitigate the consequences of an accident. Post replacement testing of the removed relays associated with PCIV CV-5065-91 and CV-5065-92 demonstrated the ability of these relays to perform the required safety function. Based on the test results, no loss of PCIS safety function occurred while the relays were physically installed and operating. Therefore, Event Number 52056, made on June 30, 2016, is being retracted. The NRC Senior Resident Inspector has been notified. Notified R1DO (Powell).

At 0238 hours (EDT) on Tuesday, August 12, 2014, with Pilgrim Station at 100 percent power in the Run Mode with reactor coolant pressure at approximately 1025 psig and the High Pressure Coolant Injection (HPCI) System previously removed from service for maintenance, a condition with the potential to impact the operability of the HPCI System was discovered. The HPCI System was being operated in accordance with plant procedures to complete post maintenance test requirements. Upon HPCI initiation, the indicated flow on HPCI Flow Indicator FI-2340-1-1 was 0 Gallons Per Minute (GPM) with the flow controller in the manual mode. The indicated flow on HPCI Flow Indicator Fl-2340-1-1 remained at 0 GPM throughout the duration of the surveillance. Alternate flow indication indicated the expected HPCI flow rate. The flow controller in manual was capable of controlling at the demanded HPCI turbine speed. The HPCI turbine speed was manually varied with a corresponding change in the HPCI flow computer point reading. Activities to restore the flow indicator capability are in progress. The plant is in a safe condition and plant personnel are investigating the cause of the flow indicator issue. The NRC Resident Inspector has been informed of this notification. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(B) and 10 CFR 50.72(b)(3)(v)(D). The licensee will be notifying the state.

• * * RETRACTION FROM O'ROURKE TO KLCO ON 10/03/2014 AT 1254 EDT * * *

Subsequent investigation determined that HPCI Flow Instrument SQRT-2340-10 output signal was 0 mA and did not change in response to the actual HPCI flow rate. With the SQRT-2340-10 output signal at 0 mA, the HPCI Flow controller would demand maximum HPCI injection flow in the AUTOMATIC control mode. Circuitry within the control system limits the maximum HPCI flow to 5250 GPM at a turbine speed of 4165 RPM. Engineering analysis has concluded that the HPCI pump operating limits (net positive suction head and low pressure suction trip) would not be exceeded in a maximum HPCI flow state. Therefore, the HPCI System was operable and capable of performing its residual heat removal and accident mitigation functions. Therefore, the initial 50.72(b)(3)(v)(B) and 10CFR50.72(b)(3)(v)(D) report is being retracted. The (NRC) Resident Inspector has been informed of this notification retraction. Notified the R1DO (Krohn).

On January 5, 2011, at 0120 hours, with the reactor at 100% thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNSP) declared the Reactor Core Isolation Cooling (RCIC) system inoperable due to the RCIC suction isolation valve from the Torus/Suppression Pool (RCIC-26) failing to go fully closed during planned surveillance testing. The RCIC-26 is a motor-operated valve (MOV) and its normal position is closed. The RClC-26 valve is redundant to the RCIC-25 valve, and is not the credited containment isolation valve. The RCIC-26 valve has a safety function to be (manually) opened during certain event mitigation scenarios requiring a transfer of suction sources from the Condensate Storage Tank (CST) to the Torus. Based on the valve failing to fully close during MOV stroke time testing per PNPS Procedure 8.5.5.4, the RCIC system was declared inoperable at 0120 hours and the appropriate LCO was entered. The RCIC-26 was subsequently returned to a full open position, caution tagged and the RCIC system was declared operable. The LCO was exited at 0200 hours. An investigation of the event is underway and continuing. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector is on-site and has been notified. This is an 8-hour notification made in accordance with 50.72(b)(3)(v)(D). The licensee will notify the State of Massachusetts.

• * * RETRACTION FROM JOSEPH LYNCH TO JOHN KNOKE AT 1946 EST ON 3/4/11 * * *

Event Notification 46521 was conservatively made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were satisfied pending the evaluation of RCIC System operability. On 01/05/11, at 0120 hours the RCIC System was declared inoperable due to uncertainty of RCIC System Operability when the Torus/Suppression Pool Suction Valve (RCIC-26) failed to go fully closed during planned surveillance testing. The valve was restored to the full open position and the valve was declared operable based on capability to meet the required safety function to fully open when RCIC pump suction from the suppression pool is required. The apparent cause evaluation concluded that valve failure was the result of high relay contact resistance in the closing control circuit components of the valve breaker. This failure prevented the valve from fully closing but had no affect on capability to open the valve. Surveillance testing verified that capability to open the valve was not affected. Corrective action was completed to clean or replace the control circuit relay contacts. Post work testing confirmed capability to open and close the valve. An extent of condition for similar breaker control circuit components was also performed. All relevant technical information is documented in the corrective action system. The failure observed did not affect the valve's required safety function and did not impact RCIC System operability. Thus there was no impact on nuclear safety. This event is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(D) . Event Number 46521, made on 01/05/2011, is being retracted. The licensee has notified the NRC Resident Inspector. Notified R1DO (Anthony Dimitriadis)

On October 22, 2008, at 1217 hours, with the reactor at 100% core thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNPS) conservatively declared the Reactor Core Isolation Cooling System (RCIC) inoperable in response to a concern regarding the reliability of aged capacitors that are installed in the RCIC flow controller. As background, the RCIC flow controller was calibrated and successfully tested on October 7th, 2008 as part of normal surveillance activities, however several of the capacitors installed in the controller were noted to be between 21 to 30 years of age. Industry recommended replacement interval for the capacitors is typically between 7 to 10 years of age. PNPS engineering review in conjunction with Entergy fleet consultation concluded today (10/22) that there was no definitive technical bases to provide a reasonable expectation that the RCIC flow controller function can be assured throughout it's mission time due to the capacitor aging concern. Therefore, RCIC was declared inoperable and a 14 day limiting condition for operability action statement was entered in accordance with TS 3.5.D.1. A replacement controller is being prepared for installation, with post maintenance testing projected to be completed by 2100 hours this evening. Ultimately the suspect controller will be the subject of further evaluation and this notification will be updated as appropriate. This notification has no impact on the health and safety of the public. The NRC Senior Resident Inspector is onsite and has been notified. This is an 8 hour notification made in accordance with 50.72(b)(3)(v)(D).

• * * RETRACTION AT 1435 EST ON 12/12/2008 FROM JOHN WHALEY TO DONALD NORWOOD * * *

Basis for Retraction: Event Notification 44593 was conservatively made to ensure that the eight-hour non-emergency reporting requirements of 10 CFR 50.72 were satisfied pending the evaluation of RCIC System operability. On 10/22/08, RCIC flow controller FIC-1340-1 was declared inoperable due to engineering uncertainty for controller operability. The controller's electrolytic capacitors appeared to be aged beyond the expected useful life, and the resultant degrading power supply voltage indicated that the controller may not operate for the required FSAR mission time of eight hours. The controller was replaced on 10/23/08 with a refurbished controller and subsequent post-maintenance RCIC system flow testing demonstrated RCIC system operability. The controller that was removed from service was evaluated. Controller bench testing was performed on 11/6 and 11/7, 2008. This testing demonstrated that the controller could provide a full demand output signal for a minimum of 15 continuous hours. During this testing, it was also determined that the power supply output voltage was not degrading. Based on this post-service controller testing, and the successful in-service RCIC flow controller calibration and system performance test conducted on 10/07/08, the controller was operable when installed. The RCIC system was capable of performing its intended safety functions and would have started and supplied design basis flow to the reactor vessel under design basis conditions. Thus there would have (been) no impact on nuclear safety. Therefore, this event was not reportable pursuant to 10CFR50.72(b)(3)(v)(D). Event Number 44593, made on 10/22/2008, is being retracted. The licensee notified the NRC Resident Inspector. Notified R1DO (Bellamy).

On November 20, 2007 at 0630 hours, with the reactor at 100% core thermal power, a power supply failure was discovered in the high pressure coolant injection (HPCI) flow controller circuitry that may have precluded the system from performing its design basis function. Therefore, in accordance with 10 CFR Part 50.72(b)(3)(v) an eight-hour notification is being made. As background, on November 18, 2007, at 2145 hours, the high pressure coolant injection (HPCI) system was removed from service for planned maintenance. The required risk analysis was performed and the appropriate 14 day limiting condition for operation (LCO) was entered in accordance with Technical Specification (TS) 3.5.C. Later on November 19, 2007 at approximately 2100 hours the planned maintenance had been completed and HPCI was restored to the normal standby line-up in preparation for post maintenance testing (PMT). The HPCI valve quarterly operability and HPCI pump and valve quarterly operability tests were performed as the prescribed PMT. Upon initiation, the HPCI turbine was observed to come up to expected rated speed (~4,200 rpm) and expected HPCI pump discharge pressure (~1,300 psig). However HPCI pump indicated discharge flow was observed to be ~2,300 gpm, which is less than the Technical Specification requirement of 4,250 gpm. The HPCI system was secured and remained in the original TS 3.5.C LCO and a troubleshooting plan was initiated. On November 20, 2007, at 0630 hours, troubleshooting identified a power supply failure in the HPCI flow control circuitry. A replacement flow controller was identified and installed and it is anticipated that appropriate PMT will be initiated by 1600 hours. The impact of the power supply failure for the design basis operability for HPCl could not be definitively established before the eight-hour notification requirement of 10 CFR Part 50.72(b)(3)(v) was exceeded. The licensee notified the NRC Resident Inspector and the Commonwealth of Massachusetts.

• * * RETRACTION FROM DAVE NOYES TO JOE O'HARA AT 1751 ON 1/14/08 * * *

NRC Notification 43794 was conservatively made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were met pending the evaluation of an atypical condition (low reading) observed with the High Pressure Coolant Injection (HPCI) Flow Controller while performing scheduled surveillance testing for the HPCI System. During surveillance testing on 11/18/07, the HPCI System was started and met or exceeded the Technical Specification minimum requirements designed to demonstrate HPCI System Operability. While testing the specific components of the system, the HPCI Flow Controller was observed to be behaving erratically. Although the HPCI System was still capable of performing its required design safety function, the Shift Manager declared the system inoperable since he did not have definitive indication that the turbine was providing the required flow. Troubleshooting of the flow controller determined that the low flow indication was due to a degraded transmitter power supply located internal to flow controller FIC-2340-1. FIC-2340-1 is located in the main control room and is used to control HPCI system flow rate, and provide power to flow transmitter FT-2358. Although indicated flow rate was only 2300 gpm due to the degraded power supply, actual flow rate was approximately 5400 gpm based on pump hydraulic curves. The power supply in question only supplies power to FT-2358. Normal required supply voltage from this power supply is 28VDC to 36VDC. The degraded power supply could only supply 22.4VDC at the transmitter FT-2358 terminals. The degraded power supply voltage caused transmitter to output a lower than normal current for the actual measured flow rate giving a false low flow rate to FIC-2340-1. An Apparent Cause Evaluation and Past Operability Evaluation were performed in response to this event. These evaluations concluded that HPCI System was capable of performing its intended safety functions with the transmitter power supply degraded. HPCI system was capable of performing its intended safety functions during the time when FIC-2340-1 transmitter power supply exhibited low output voltage. HPCI would have started and supplied design basis flow to reactor vessel under design basis conditions. Thus there would have no impact on nuclear safety. Therefore, this event was not reportable pursuant to 10 CFR 50.72(b)(3)(v)(D). ENS Event Number 43794, made on 11/20/2007, is being retracted. The licensee will notify the NRC Resident Inspector and the Massachusetts Civil Defense Authority. Notified R1DO(Cobey)

The report is being made in accordance with 10 CFR 50.72(b)(3) due to both Emergency Diesel Generators (EDGs) being declared inoperable. (This is a 24-hour LCO.) The EDGs were declared inoperable due to indicated outside air temperature exceeding the established operability limit of 95F for two short periods of time. Indicated outside air temperature went above 95F during the following time frames: 10:55 to 11:00 and 12:15 to 12:20. Indicated outside air temperature subsequently decreased to less than 95F following these spikes. Both EDGs are currently operable. Outside ambient temperatures are continuing to be monitored. Offsite power is available. Further evaluation of this event is ongoing. The licensee notified the NRC Resident Inspector.

• * * EVENT RETRACTION FROM LICENSEE (NOYES) TO ABRAMOVITZ AT 12:06 ON 08/12/2005 * * *

The initial report was made in accordance with 10 CFR 50.72(b)(3) due to both Emergency Diesel Generators (EDGs) being declared inoperable. The EDGs were declared inoperable due to indicated outside air temperature exceeding the established operability limit of 95F for short periods of time on 6/26/05. Outside indicated air temperature subsequently decreased to less than 95F following these short timeframes. Both EDGs were operable when the notification was made, and outside ambient temperatures continued to be monitored. Offsite power was available. Further evaluation of this event has been performed. The initial report used temperature indication that was conservative with respect to actual conditions. The evaluation determined that the air temperature outside the EDG Building was a maximum of 91 F on 6/26105. This evaluation was based on other temperature reading taken during the periods of interest. Additionally, further evaluation has determined that both EDGs were not inoperable even at the higher temperatures initially indicated. Therefore, the EDGs were not inoperable during the periods reported and event notification #41799 is retracted. Notified the R1DO (Caruso). The licensee notified the NRC Resident Inspector.