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EN Revision Text: POTENTIAL LOSS OF MSIV SCRAM FUNCTION DURING MAIN STEAM LINE ISOLATION VALVE TESTING At approximately 1040 EST on January 5, 2019, during evaluation of test results for the 'C' Main Steam Isolation Valve (MSIV), it was determined that closure of three of four Main Steam Lines would not necessarily have resulted in a full scram during testing due to failure of a limit switch (LS-6) associated with MSIV-1C while in the test configuration. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition.' The system was restored from the testing configuration at 1057 EST and the failed trip channel was placed in the tripped condition at 1326 EST thus restoring the design function. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

• * * RETRACTION AT 1529 EST ON 02/11/19 FROM JOSEPH FRATTASIO TO JEFF HERRERA * * *

The purpose of the notification is to retract ENS Notification 53815 made on 01/05/19 for Pilgrim Nuclear Power Station. The previous notification reported that there was a potential loss of Main Steam Isolation Valve (MSIV) scram function during main steam line isolation valve testing, at the time of discovery, due to failure of a limit switch (LS-6) associated with MSIV-1C while in the test configuration. Subsequent evaluation has demonstrated that the scram function credited in the design basis was not lost. Specifically, after an Engineering Evaluation, it has been determined that the MSIV position RPS logic was not lost for those functions within the design basis and, as such, was capable of performing its intended safety function. The NRC Resident Inspector has been notified. Notified the R1DO (Cahill).

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).

On Sunday, April 14, 2013 at 2216 hours, with the Pilgrim Nuclear Power Station (PNPS) Reactor Mode Select Switch (RMSS) in Start-up, the turbine generator previously removed from service, and the reactor sub-critical on Intermediate Range Monitors Range 2 and lowering, the PNPS Containment Personnel Air Lock failed integrated air lock testing as required by TS 4.7.A.2. 10CFR50 Appendix J requires that primary reactor containment meet certain leakage rate testing requirements. These test requirements ensure that 1) Leakage through the containment or systems and components penetrating the containment do not exceed allowable leakage rates specified in Technical Specifications and 2) The integrity of the containment structure is maintained during its service life. The test requirements include local leakage rate testing of containment air locks. The test criteria establishes a limit of less than or equal to 10.525 SLM, actual leakage was 16.7 SLM. PNPS was in the process of shutting down for a scheduled Refueling Outage during the scheduled testing. This event had no impact on the health and/or safety of the public. The USNRC Resident Inspector will be notified. This 8-hour notification is being made in accordance with 10CFR50.72(b)(3)(v)(c).

* * * RETRACTION FROM BOB O'NEILL TO PETE SNYDER AT 1700 ON 6/13/13 * * * 

Subsequent investigation of the test failure determined that an o-ring in the Airlock Inner Pressure Equalizing Device (PED) was not properly seated resulting in exceeding the established leakage rate limit. Leakage was not observed on the Airlock Outer Boundary Surface. Thus, with the Airlock Outer Boundary Surface intact, the Airlock was capable of performing the safety function to control the release of radioactive material. In addition, even with the Airlock leakage of 16.7 standard liters per minute (SLM), the Technical Specification Primary Containment As-Found Minimum Pathway Leakage Limit of 0.6 La for all Type B and C leakage tests was not exceeded. Therefore, the initial 50.72(b)(3)(v)(C) report is being retracted. The licensee will notify the NRC Resident Inspector. Notified R1DO (Dentel).

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).

This report is being made in accordance with 10CFR50.72 (b) (3) (v) due to the High Pressure Coolant Injection (HPCI) system being declared inoperable. HPCI was declared inoperable on 8/25/05 at 1630 EST due to oscillations at below rated flow during the scheduled operability testing. HPCI was restored to standby line-up when testing was completed and remains available for use. This event is an eight hour notification. Efforts are on going to determine the cause of the oscillations on the Flow Controller. This event had no adverse effect to the health and safety of the public. The resident NRC inspector has been notified of this event.

• * * RETRACTION FROM D. NOYES TO W. GOTT AT 1711 ON 09/27/05 * * *

This follow-up notification is being made to retract the notification made to the NRC Operations Center on 8/25/05 at 2039 hours (notification #41948). The initial report was made in accordance with 10 CFR 50.72(b)(3) due to the HPCI system being declared inoperable. The system was declared inoperable due to oscillations in turbine speed, pump discharge pressure, and pump flow during a quarterly surveillance test of the HPCI pump. Further investigation and evaluation of this has been performed. The cause of the noted oscillations was the position of a hand operated valve that is located in the HPCI system full flow test line. The hand operated valve is located downstream of an in-series motor operated valve that automatically closes if an automatic HPCI system initiation signal occurs. The full flow test line is not part of the HPCI injection pathway to the reactor vessel. As a result, the position of this valve would not have impacted the ability of the system to perform its design function. After adjusting the position of the hand operated valve, the surveillance test of the HPCI pump was completed with satisfactory results. The evaluation has determined that the HPCI system was capable of performing the designed safety function. Therefore, the HPCI system was not inoperable and event notification #41948 is retracted. The licensee notified the NRC Resident Inspector. Notified R1DO (C. Cahill)