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 Report dateSiteEvent description
05000293/LER-2017-01325 January 2018Pilgrim

On November 26, 2017, with the Reactor in the Run Mode at 100 percent power, while reviewing a procedure to be performed during normal scheduled testing it was determined that the test as written would cause both trains of Standby Gas Treatment System (SGTS) to be made inoperable during the test. This also made secondary containment system (SCS) inoperable. This LER is submitted to acknowledge that Pilgrim Nuclear Power Station missed providing Event Notifications and LERs for past occurrences. With both trains of SGTS and SCS inoperable while in Run, this event is reportable in accordance with Title 10 Code of Federal Regulations 50.73(a)(2)(v)(C) and 50.73(a)(2)(v)(D) as conditions that could have prevented the fulfillment of the safety function of a structure or system needed to control the release of radioactive material and mitigate the consequences of an accident. This has been determined to be a reportable condition that has not been reported during the past three years involving SGTS and secondary containment inoperability. The reportable conditions have occurred several times within the past three years during scheduled testing of SGTS.

This event had no impact on the health and/or safety of the public.

05000293/LER-2017-00915 November 2017Pilgrim

On May 17, 2017, during Refueling Outage (RFO)-21 while performing an extent of condition review it was discovered that the contact indicating tabs of relays 16A-K30 and 16A-K54 of the Pilgrim Nuclear Power Station (PNPS) Primary Containment System, were visually hanging in the mid-position (partial travel).

The relays were replaced during RFO-21 along with 16A-K29 and 16A-K53, and all four relays were sent to an offsite vendor for further testing and analysis. Other relays were reviewed but were determined to be outside the scope of this extent of condition review.

PNPS stated at the time that this event was reportable under 10 CFR 50.73(a)(2)(i)(B) - Operation or condition prohibited by Technical Specifications; and potentially reportable in accordance with 10 CFR 50.73(a)(2)(v)(B), (C) and (D) - Any condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat, control the release of radioactive material and mitigate the consequences of an accident. However, additional information provided by our offsite vendor and an engineering evaluation, support the conclusion that there was never a loss of safety function regarding any of the four relays (16A-K29, 16A- K30, 16A-K53 and 16A-K54). Therefore, this event was not reportable under 10 CFR 50.73(a)(2)(i)(B) nor under 10 CFR 50.73(a)(2)(v)(B), (C) or (D).

This event posed no threat to public health and safety.

05000293/LER-2017-00313 November 2017Pilgrim

operators were restoring a valve lineup after performing planned operations when a pathway was created through valve manipulation that allowed water in the Condensate Storage Tank to be diverted to the suppression pool. The rise in suppression pool water level exceeded the maximum water level specified in Technical Specifications (TS) and caused a lowering of the drywell to suppression pool differential pressure.

With suppression pool water level and differential pressure both outside of TS limits, Limiting Condition for Operation Action Statements 3.7.A.5 and 3.7.A.8.c were both entered and investigation into the cause of diverting water into the suppression pool was initiated. Water level was restored to within TS limits at 1540 (EDT).

An 8-hour non-emergency notification was made in accordance with 10 CFR 50.72(b)(3)(v)(D), any event or condition that at the time discovery had the potential to have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

The initial LER was submitted under 10 CFR 50.73(a)(2)(i)(B), operation or condition prohibited by TS, but it was subsequently determined it should have been submitted under 10 CFR 50.73(a)(2)(v)(D), any event or condition that at the time of discovery had the potential to have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident, to more precisely describe this condition.

There was no impact to public health and safety from this condition.

05000293/LER-2017-01213 November 2017Pilgrim

On September 13, 2017 with the unit at 100 percent power, Pilgrim Nuclear Power Station discovered during calibration of four Start-Up Transformer Degraded Voltage Relays (ABB model number ITE- 27N) after the relays had been removed from the plant, relay 127A-604/1 had an as-found setpoint value outside the Technical Specifications Table 3.2.B limit of 110.6 plus or minus 0.56 Volts ac (Vac) (110.04 - 111.16); the as-found value was 109.90 Vac. This relay was replaced with a spare relay that was calibrated prior to its installation.

Pilgrim Nuclear Power Station is submitting this Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(i)(B) - Any operation or condition that was prohibited by the plant's Technical Specifications.

This event was not risk significant. There was no threat to public health and safety from this condition.

05000293/LER-2017-0082 November 2017Pilgrim

On May 3, 2017 with the unit shutdown for refueling outage, while performing plant procedure 3.M.3-27, "480V Bus B6 Transfer Test, UV, Degraded Voltage and Timing Relays Calibration and Annunciator Verification," the time delay Agastat relay 27A-B1X/TDDO opened instantaneously, instead of with a time delay. This relay is set to drop out after a 1.25 second time delay after being de-energized. This condition was discovered during the plant's refueling outage when conditions were such that the equipment normally energized/activated by this time delay relay was not required to be operable.

Pilgrim Nuclear Power Station is submitting this Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(i)(B) - Operation or condition prohibited by Technical Specifications; and in accordance with10 CFR 50.73(a)(2)(v)(B), (C) and (D) - Any condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat, control the release of radioactive material and mitigate the consequences of an accident.

This event was not risk significant. There was no threat to public health and safety from this condition.

05000293/LER-2017-01115 August 2017A Ler Number
Pilgrim

On June 20, 2017, at 1444 hours (EDT), with the reactor at 100% core thermal power and steady state conditions, plant personnel notified the Main Control Room that both doors in the secondary containment airlock at the 23 foot elevation on the East Side Reactor Building (RB) Entrance were opened simultaneously.

The failure of this interlock (to prevent both doors from being opened) caused a loss of secondary containment per Technical Specification (TS) 3.7.C.1. The doors were immediately closed, and the secondary containment boundary was reestablished.

An 8-hour non-emergency notification was made in accordance with 10 CFR 50.72(b)(3)(v)(C), any event or condition that at the time discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. The safety significance of this event was minimal given the impact on the secondary containment.

Secondary containment remained available and functional during the event since secondary containment was immediately restored by closing the doors.

05000293/LER-2017-0103 August 2017Pilgrim

On June 6, 2017, at 1357 (EDT) with the reactor at 100% core thermal power and steady state conditions, plant personnel were performing Ultrasonic testing (UT) examination on Core Spray A high point piping in the A Residual Heat Removal Quad to ensure this piping was water solid, when they identified that the top of this horizontal pipe had an air void internally. A high point vent line with valves is located in the area of the UT exam.

It was found that the top 2 inches of the 10 inch core spray pump discharge line had accumulated an air void within the known inverted loop in Core Spray Loop-A discharge line. The system had been drained for maintenance during the Refueling Outage.

Pilgrim Nuclear Power Station is reporting this event pursuant to 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by Technical Specifications. Although upon discovery the proper Limiting Condition for Operation Action Statement was entered and the void was filled immediately to correct the issue, it is believed that this condition existed before the time of discovery for a period of time longer than that allowed by Technical Specifications.

This event posed no threat to public health and safety.

05000293/LER-2017-00117 July 2017Pilgrim

Station (PNPS) was performing surveillance testing of secondary containment isolation dampers when dampers AO-N-82 and AO-N-83, refueling floor supply isolation dampers, failed to fully close when the control switches were taken to close.

The failure of dampers AO-N-82 and AO-N-83 to fully close resulted in a loss of safety function for secondary containment, causing immediate entry into Limiting Condition for Operation (LCO) Action Statement (AS) 3.7.C.2.a, at 1155 hours. This LCO AS was exited at 1206 hours when the dampers were verified closed.

An 8-hour non-emergency notification was made in accordance with 10 CFR 50.72(b)(3)(v), any event or condition that at the time discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. In addition, this notification is being conservatively made by PNPS in accordance with 10 CFR 50.73(a)(2)(i)(B), as a condition that was prohibited by Technical Specifications.

The reactor building isolation dampers were cleaned, lubricated and post-work tested. PNPS has returned the dampers to operable status. Planned action to prevent recurrence is to revise the preventive maintenance strategy.

There was no impact to public health and safety from this condition.

05000293/LER-2017-00613 June 2017Pilgrim

On April 15, 2017, with the reactor at 0 percent core thermal power in a refueling outage, Pilgrim Nuclear Power Station discovered that during fuel movements (core alterations) conducted on April 14th and April 15th in reactor quadrant 'B', source range monitor channel B (SRM B) was inoperable due to not meeting the signal to noise ratio of no less than 2 to 1 required by FSAR section 7.5.4.1.

This event is reportable under 10 CFR 50.73(a)(2)(i)(B), any operation or condition which was prohibited by the plant's Technical Specifications.

There was no impact to public health and safety from this condition.

05000293/LER-2017-0057 June 2017Pilgrim

On April 10, 2017, the Personnel Airlock, X-2, failed to meet local leak rate test acceptance criteria. This failure is reportable under 10 CFR 50.73(a)(2)(i)(B), any operation or condition which was prohibited by the plant's Technical Specifications.

On April 22, 2017, the High Pressure Coolant Injection System turbine exhaust line check valves both failed to meet local leak rate test acceptance criteria. The test volume for each valve could not be pressurized when flow was greater than 100 Standard Liters per Minute. Significant air flow was coming out of the test vent, indicating that each check valve was either degraded or not seated.

This failure resulted in the current Refueling Outage summation of Type B and Type C testing results exceeding the 10 CFR 50, Appendix J local leak rate test criteria limit of 0.6 La and the primary containment total leakage criteria limit of 1.0 La. This created an event that is reportable under 10 CFR 50.73(a)(2)(i)(B), any operation or condition which was prohibited by the plant's Technical Specifications,10 CFR 50.73(a)(2)(ii)(A), any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded, 10 CFR 50.73(a)(2)(v)(C), any condition that could have prevented the fulfillment of a safety function of a system needed to control the release of radioactive material, and 10 CFR 50.73(a)(2)(v)(D), any condition that could have prevented the fulfillment of a safety function of a system needed to mitigate the consequences of an accident.

There was no impact to public health and safety from this condition.

05000293/LER-2017-0042 June 2017Pilgrim

On April 5, 2017, at 0030 hours (EDT) with the Reactor in the Run Mode at approximately 97 percent power, both trains of the Standby Gas Treatment System (SBGTS) were made inoperable during the performance of a surveillance test of secondary containment prior to the refueling outage, With both trains of SBGTS inoperable while in the Run mode, this event is reportable per the requirements of Title 10, Code of Federal Regulations (CFR) 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D), any event that could have prevented the fulfillment of the safety functions to "control the release of radioactive material" and "mitigate the consequences of an accident.

This event had no impact on the health and/or safety of the public.

05000293/LER-2017-00225 May 2017Pilgrim

On March 27, 2017, at 1825 (EDT), with the reactor at 100 percent core thermal power and steady state conditions, plant personnel caused a High Pressure Coolant Injection (HPCI) System isolation. Pilgrim Nuclear Power Station was performing planned testing on the Reactor Core Isolation Cooling (RCIC) when the HPCI System isolated. Accordingly, the HPCI System was declared inoperable.

The Technical Specifications Limiting Condition for Operation 'Action Statement 3.5.C.2 was entered and planned troubleshooting into the cause of the HPCI isolation was started. This event is reportable under 10 CFR 50.73(a)(2)(v)(D), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

There was no impact to public health and safety from this condition.

05000293/LER-2016-01013 February 2017Pilgrim

On December 15, 2016, at 1500 (EST), with the reactor at approximately 22 percent power, the Main Steam Isolation Valves (MSIVs) 2C and 2D were discovered to have steam leaks while performing a steam tunnel walkdown. MSIV 2D, which had a body to bonnet steam leak, was declared inoperable and Technical Specifications (TS) Limiting Condition for Operation Action Statement (LCOAS) 3.7.A.2.b was entered at 1530 on December 15, 2016. Outboard MSIV 2D and inboard MSIV

  • 1D both were closed and deactivated to isolate Main Steam Line D. On December 16, 2016, at 1524 (EST) Operations entered TS LCOAS 3.7.A.2.b for outboard MSIV 2C. Actions were also taken to close and deactivate the inboard MSIV 1C, which included a controlled plant shutdown to reduce reactor pressure below the MSIV closure scram bypass setpoint.

Based on the evidence found, it was reasonable to conclude that the MSIV 2D valve body to bonnet steam leak and the MSIV 2C packing leak had likely started sometime prior to the event date and both were leaking for a period of time greater than that allowed by TS. Therefore, PNPS is making this submittal in accordance with 10 CFR 50.73(a)(2)(i)(B), any operation or condition prohibited by the plant's TS. In addition, PNPS closed the inboard MSIV 1C in accordance with TS LCOAS 3.7.A.2.b prior to going to Cold Shutdown. However, PNPS is also conservatively making this submittal in accordance with 10 CFR 50.73(a)(2)(i)(A), the completion of any nuclear plant shutdown required by the plant's TS.

The plant was placed in Cold Shutdown and both the outboard MSIV 2C and 2D were repaired and returned to service.

There was no impact to public health and safety from this condition.

05000293/LER-2016-0096 January 2017Pilgrim

On November 7, 2016, at 1609 (EST), with the reactor at 100 percent core thermal power and steady state conditions, the High Pressure Coolant Injection (HPCI) system was declared inoperable. Pilgrim Nuclear Power Station was performing planned quarterly testing per Technical Specifications 4.13.A.1. During a review of the HPCI pump data taken during the test, it was determined that the recorded vibration reading on the main pump outboard horizontal point (P4H) was 0.8335 in/sec which exceeds the 1ST required action range high limit of less than or equal to 0.830 in/sec. Accordingly, the HPCI pump was declared inoperable.

The Limiting Condition for Operation Action Statement 3.5.C.2 was entered and planned troubleshooting into the cause of the high vibration was started. This event is reportable under 10 CFR 50.73(a)(2)(v)(D), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

There was no impact to public health and safety from this condition. ,

05000293/LER-2016-0089 December 2016Pilgrim

On September 28, 2016, while performing the pre-start checks prior to running the Emergency Diesel Generator (EDG)-A for the monthly Technical Specification (TS) surveillance, the oil level in the EDG radiator fan right angle gearbox was found to be low and additional checks found the gearbox oil pressure relief valve in a loose state which provided a pathway for gear oil to be pumped out of the gear box while the EDG was operating. EDG-A was declared inoperable, the relief valve was repaired, pressure tested and the pressure adjusting threaded union was staked to eliminate any risk from vibration induced motion in the future, the gearbox oil was replaced and the EDG run for a post-maintenance test.

A Functional Failure Determination completed on October 11, 2016 determined that the EDG would not have been able to run for its stated mission time of 30 days. This condition existed for a period of 28 days since the last surveillance test on August 31, 2016 which is greater than the TS Allowed Out of Service Time (AOT) of 72 hours. However, the Station Black Out Diesel Generator was available during this time frame. This issue is reportable under 10 CFR 50.73(a)(2)(i)(B) as an Operation or Condition which was Prohibited by the plant's TSs.

On September 15, 2016 EDG-B was made inoperable to perform its monthly operability run. This created a situation where for a brief period of time both EDGs were inoperable which is a condition that could have prevented the fulfillment of the safety function of a system needed to shut down the reactor and maintain it in a safe condition, remove residual heat, and mitigate the consequences of an accident which is reportable in accordance with 50.73(a)(2)(v)(A), 50.73(a)(2)(v)(B), and 50.73(a)(2)(v)(D). EDG-B remained available and could quickly have been restored by manual action to an operable condition if needed during the operability run.

There was no impact to public health and safety from this condition.

05000293/LER-2016-00317 November 2016Pilgrim

On May 12, 2016, at 16:47 EDT, an assessment of Boron Areal Density Gauge for Evaluating Racks (BADGER) testing results for spent fuel storage racks revealed a single panel, RR35 South with a gap in neutron absorber material that exceeds spent fuel storage design feature assumptions. The assumptions ensure compliance with Technical Specification (TS) Section 4.3.1.1.b to maintain a spent fuel pool K-effective (Kell) of less than or equal to ( that may further challenge compliance with TS 4.3.1.1.b.

The cause of the noncompliance with TS 4.3.1.1.b is Boraflex degradation greater than assumed in the criticality safety analysis. Interim actions include declaring the spent fuel pool rack cells adjacent to the panel RR35 South inoperable, installing blade guides into cells adjacent to panel RR35S, and prohibiting fuel movement in Boraflex racks. An additional corrective action was taken to perform a criticality evaluation of the spent fuel pool (SPF).

This LER supplement is issued to identify that an additional criticality evaluation has been performed to conservatively identify fuel storage configurations within the Boraflex fuel storage racks that will ensure the Boraflex racks Keff will be able to be maintained configurations.

05000293/LER-2016-0074 November 2016Pilgrim

On September 6, 2016 at 0827 EDT, with the reactor at approximately 91 percent core thermal power, operators manually scrammed the reactor when the benchmark for a reactor water level of +42 inches increasing was reached. Following the scram, all rods fully inserted and the Average Power Range Monitors were downscale, indicating the reactor was shut down. The Main Steam Isolation Valves closed on a Primary Containment Isolation Signal Group 1 isolation and were subsequently reopened to maintain reactor pressure.

The reason for the increasing reactor water level was the malfunction of Feedwater Regulating Valve 'A' (FRV 'A'). The reactor operator at the control panel placed FRV 'A' in remote manual in an attempt to stabilize the feedwater flow oscillations. This had no effect on the performance of FRV 'A'. The operators experienced feedwater flow oscillations from FRV 'A' that resulted in reactor water level high and low level alarms on the control panel. Shortly after this, the benchmark reactor water level of +42 inches increasing was reached and operators manually scrammed the reactor.

There was no impact to public health and safety.

05000293/LER-2016-00513 October 2016Pilgrim

Nuclear Power Station (PNPS) declared the ultimate heat sink (UHS) and salt service water (SSW) system inoperable due to high sea water inlet temperatures greater than 75 degrees Fahrenheit (F). PNPS had already taken action, in accordance with plant procedures, to reduce power from 100 percent in an effort to keep from exceeding the Technical Specification (TS) Limit. PNPS entered a 24-hour shutdown Limiting Condition for Operation Action Statement (LCO-AS) for Salt Service Water (SSW) inlet temperature exceeding the TS limit in TS 3.5.6.4. The LCO-AS was subsequently exited at 1651 hours when the temperature of SSW trended to below the TS limit.

Under certain design conditions, the SSW system is required to provide cooling water to various heat exchangers such as the Reactor Building Closed Cooling Water (RBCCW) and Turbine Building Closed Cooling Water (TBCCW) systems. When the inlet temperature to these supplied loads exceeds the 75 degree F limit established in the TS, the SSW system is conservatively declared inoperable until the temperature trends below this value. This condition existed for 59 minutes reaching a maximum of 75.1 degrees F. The cause of the sea water inlet temperature exceeding the 75 degree F TS criterion was sustained increased sea water surface temperature in Cape Cod Bay due to summer weather conditions and recirculation of water from the plant's discharge due to wind and tidal conditions.

There was no impact to public health and safety from this condition.

05000293/LER-2016-00613 October 2016Pilgrim

On 8/16/2016, during performance of Main Steam Isolation Valve (MSIV) stroke time testing, with the plant operating at approximately 60% power, "C" inboard MSIV closure time was 7.4 seconds, exceeding the maximum closure time of 5 seconds referenced in Final Safety Analysis Report (FSAR) Table 5.2-4. The MSIV was declared inoperable. In accordance with the plant Technical Specifications (TS) the "C" outboard MSIV was closed and isolated. A team was formed to perform troubleshooting and determine corrective actions. With one main steam line inoperable the plant was limited to less than 75% power. On 8/21/16 the plant was shut down and entered a forced outage to repair the "C" inboard MSIV. The air pack was replaced and post maintenance tests were satisfactorily performed to restore the valve to Operable status.

On 8/30/16 Pilgrim Nuclear Power Station conservatively concluded that, based on the historical issues with the "C" inboard MSIV, the valve had been inoperable since stroke time testing was performed on May 24, 2016. i There was no impact to public health and safety from this condition.

The function of the MSIVs is to prevent coolant inventory loss and protect plant personnel in the event of a steam line break outside containment. Also MSIVs provide a primary containment boundary after a loss of coolant accident (LOCA). The MSIVs are 20-inch air/spring operated, balanced "Y"-type globe valves.

There are four main steam lines. Each steam line has two MSIVs; one inside primary containment and one outside of primary containment. The MSIV "C" inboard valve is located inside primary containment, within the drywell.

Pneumatic power to the MSIVs located inside primary containment is supplied by nitrogen when the drywell atmosphere is inerted and by instrument air when the drywell is de-inerted. An air pack consisting of two- way, three-way and four-way valves and Alternating Current/Direct Current (AC/DC) solenoid valves are used to control MSIV position (open/closed).

The air pack employs a two-stage control scheme whereby pilot air pressure controls the position of a dual- cylinder four-way valve; repositioning spools within the cylinders to apply pneumatic power above or below the MSIV air actuator piston. The air pack has one AC and one DC solenoid operated valve for controlling pilot air pressure within the air pack.

Energizing either the AC or the DC solenoid valve individually will cause the MSIV to open. Pilot air (nitrogen) passes through the energized solenoid valves to the four-way valve, repositioning its spools to port air (nitrogen) beneath the MSIV air actuator piston which drives the valve stem and main poppet open.

To close the MSIV, the AC and DC solenoids must both be de-energized to vent pilot air from the four-way valve, repositioning its spools to vent nitrogen pressure below the actuator piston and to port air (nitrogen) above the MSIV air actuator piston which drives the MSIV closed (fast closure mode). De-energizing either the AC or the DC solenoid valve individually will not close the MSIV.

The MSIV is opened and held in the open position by compressed gas (air or nitrogen). Pneumatic pressure is required to compress the springs and open the MSIV. The air actuator assists the springs during fast closure of the MSIV. The springs extend when the valve closes. If loss of air or nitrogen gas pressure to the MSIV actuator occurs then spring force exerted upon the bottom spring seat will close the MSIV (fail-safe closure). The actuator provides no air-assist during fail-safe valve closure. The valve's fail-safe closure time is slower than its air-assisted fast closure time.

During fast closure mode, the MSIV must close in the required three to five seconds elapsed time using closure force provided by both its actuator and stored energy in the compressed springs. The MSIV utilizes a dashpot to control the valve closing speed The actuator stem passes through the dashpot. A piston threaded onto the actuator stem strokes within the dashpot cylinder, displacing dashpot hydraulic fluid through a needle valve (speed control valve). MSIV closing time can be controlled between three and five seconds by adjusting the speed control valve opening (orifice).

EVENT DESCRIPTION

On 8/16/2016, with Pilgrim Nuclear Power Station operating at approximately 60% Power during performance of MSIV Operability testing, closure time for the "C" inboard MSIV was 7.4 seconds which exceeded the maximum closure time of five seconds referenced in FSAR Table 5.2-4. The "C" inboard MSIV was declared inoperable. In accordance with plant TS the "C" outboard MSIV was closed and deactivated. The plant was restricted to less than 75% power due to only having three operating main steam lines. A decision was made to shut down and replace the air pack four-way valve and solenoid valves. The drywell was entered and an as-found inspection of the MSIV air pack was completed while the MSIV was open. The air pack four-way valve and solenoid valves were replaced. The pilot air pneumatic line was cleaned and sampled for debris. The "C" inboard MSIV was tested satisfactorily and declared operable.

CAUSE OF THE EVENT

The Root Cause of this event is system debris (accumulated dust/wear and corrosion products) in the "C" inboard MSIV pilot air tubing which were disturbed during the August 2015 pipe failure and subsequently collected in the solenoid valve soft disc. The vibration and high flow through the airline resulted in accumulated system debris being agitated and released into the airline. Subsequent MSIV stroke time testing moved the debris through the air line into the newly overhauled air pack and deposited it in the solenoid valve seats, resulting in slow stoke time of the "C" inboard MSIV. The solenoid valves had been replaced in 2015 during RFO 20. At the same time, a new four-way valve had also been installed during preventive maintenance.

CORRECTIVE ACTIONS

The "C" Inboard MSIV failed its TS required stroke time test. Therefore, the plant was shut down and the air pack four-way valve and solenoid valves were replaced.

The corrective action to prevent recurrence is to replace the Instrument Air header in the drywell along with all of the pilot air lines to the inboard MSIVs. This is currently scheduled to be accomplished during RF021.

SAFETY CONSEQUENCES

There were no consequences to the safety of the public, nuclear safety, industrial safety or radiological safety due to this event.

There were no potential consequences because the "C" inboard MSIV closed as required. The delay in the timing of closure is not significant. The "C" inboard MSIV is a fail-safe design and will fail closed. In addition, if the "C" inboard MSIV failed to close, reactor isolation is available by the "C" outboard MSIV. That valve remained operable and could have performed its design function if isolation had been necessary.

Based on the defense in depth fail-safe design of inboard and outboard MSIVs, risk is considered to be Low.

No actions to reduce the frequency or consequence are necessary.

REPORTABILITY

This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Condition Prohibited By Technical Specifications.

PREVIOUS EVENTS

A review of Pilgrim Nuclear Power Station Licensee Event Reports for the past three years did not identify any similar occurrences.

REFERENCES:

Condition Report CR-PNP-2016-05987 Condition Report CR-PNP-2016-02163

05000293/LER-2016-00218 August 2016Pilgriin Nuclear Power Station
Pilgrim

, On April 19, 2016, at approximately 1450 hours, it was discovered that a maintenance activity performed between 2010 hours on August 26, 2014 and 0143 hours on August 27, 2014, had rendered the Startup Transformer (X4) and the standby Emergency Diesel Generators (EDG) (X-107A&B) unable to automatically supply power to Buses A5 and A6, due to the breaker interlock that would prevent Startup Transformer breakers (152-504 and 152-604) and standby EDG breakers (152-509 and 152-609) from closing, when Bus A8 to Bus A5 breaker (152-501) and Bus A8 to Bus A6 breaker (52-601) are in the TEST position and CLOSED. During the maintenance activity, the plant was operating at 100 percent power and the_ Unit Auxiliary Transformer (X3) was providing power to Emergency Buses A5/A6.

The functional testing of negative sequence relays (146-600/A and B) and 23kV feed undervoltage relays (127-600A/1 and 2, and 127-600B/1 and 2) created a test configuration, lasting less than 1-hour, whereby power to Buses A5 and A6 was not automatically available from either the startup transformer or from the EDGs. As a result, Limiting Conditions for Operation (LCO) Action Statement 3.9.6.2 was not met.

The root cause is that the decision to perform the described surveillance testing online, instead of during cold shutdown, lacked sufficient rigor to ensure compliance with Technical Spepifications. Corrective actions will establish and institutionalize expectations and accountability for station leadership regarding consequence-biased decision-making and effective risk manaaement. There was no impact to_public health and safety.

05000293/LER-2016-00411 July 2016Pilgrim

past operability of the Salt Service Water (SSW) Pump P-208B power supply breaker overload relay based on thermography results obtained on October 25, 2013. A similar high temperature condition was identified during the February 29, 2016 thermography surveillance that resulted in declaring the pump inoperable on March 4, 2016 and completing the replacement of the overload relay on March 5, 2016. SSW Pump P-208B is currently operable.

The past operability review covers the period of time between the 2013 thermography surveillance and March 5, 2016 when the pump breaker overload relay was replaced. Based on the lack of detailed information in the 2013 surveillance, past operability of the pump could not be assured. A review of out-of-service time for SSW Pumps P208A and P208C indicates that one instance occurred where less than two SSW pumps were Operable to feed the SSW A Loop Header for a period in excess of SSW Technical Specification requirements. This represents a reportable condition.

The cause of the overload relay high temperature readings was evaluated and determined to be a high resistance connection on the overload relay heater assembly. Inadequate predictive maintenance monitoring was identified as the apparent cause for not periodically assessing breaker status or replacing the breaker overload relay prior to March 5, 2016. Testing the overload relay that was removed from service is planned. This testing is expected to be completed by, ..

August 15, 2016. Results will be reported in a supplement to this report.

There was no impact to public health and safety.

05000293/LER-2016-0019 June 2016Pilgrim

On April 12, 2016 at 0050 hours, with the reactor at 100% power and the mode switch in RUN, Pilgrim Nuclear Power Station (PNPS) entered an unplanned 24-hour Limitihg Condition for Operation (LCO) Action Statement per Technical Specification 3.5.F.1 due to both emergency diesel generators (EDG) being inoperable. With EDG B out of service for a planned LCO maintenance window, EDG A was declared inoperable due to identification of an approximate leak of 130 drops per minute (dpm) from the line to the jacket water pressure indication.

The apparent cause of the EDG A jacket water pressure boundary leak was stress corrosion cracking (SCC) that eventually led to failure of a bulkhead pipe fitting. The apparent cause was confirmed by the failure analysis performed. Corrective action was taken to replace the leaking bulkhead pipe fitting on EDG A to restore EDG Operability. An aditional corrective action is planned to replace the opposite train EDG jacket water bulkhead pipe fitting.

There was no impact to public health and safety.

05000293/LER-2015-00212 May 2015Pilgrim

On March 12, 2015, after further evaluation of system performance of SRV-3A and SRV-3C, along with results of valve internal conditions identified during physical inspection, the valves were determined to have been inoperable for an indeterminate period during the last operating cycle. Specifically, SRV-3C was determined to be inoperable based on its on-demand performance at low reactor pressures, as well as the visual conditions that were identified during the inspection process. SRV-3A was considered inoperable based on it having similar internal indications as SRV-C when it was disassembled and inspected. SRV-3A Was installed in May 2011 and SRV-3C was installed in October 2013.

Additionally, during an extent of condition review of historical SRV performance, the review identified on March 13, 2015 that SRV-3A had failed to open in response to three manual actuation demands on February 9, 2013.

At the time the valves were declared inoperable the reactor was at 100% power. The valves had been replaced in February 2015 during the forced outage relating to winter storm Juno. This event posed no threat to public health and safety.

05000293/LER-2015-00130 March 2015Pilgrim

On Tuesday January 27, 2015, at 0402 hours, while in the process of lowering reactor power, with the reactor in the RUN mode at 52 percent core thermal power, Pilgrim Nuclear Power Station (PNPS) experienced a loss of 345KV power resulting in a load reject and an automatic reactor scram. The loss of 345KV power was due to faults from flashovers in the PNPS switchyard. All control rods fully inserted.

The Emergency Diesel Generators had been previously started and were powering safety-related buses A5 and A6. The plant stabilized in Hot Shutdown. At the time of the event a significant winter storm (Juno) was buffeting Southern New England.

The root cause of the event is that the design of the PNPS switchyard does not prevent flashover when impacted by certain weather conditions experienced during severe winter storms. A modification of the switchyard is planned to address the susceptibility of the PNPS switchyard to flashovers during severe winter storms.

This event posed no threat to public health and safety.

05000293/LER-2014-0017 July 2014Pilgrim

On May 6, 2014, with Pilgrim Nuclear Power Station (PNPS) in the RUN Mode operating at 100 percent power, the NRC Resident Inspector raised a concern about the PNPS method of complying with PNPS Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.A.2.b when a Primary Containment Isolation Valve is inoperable. TS LCO 3.7.A.2.b.requires that at least one containment isolation valve in each line having an inoperable valve shall be deactivated in the isolated condition. Deactivation means to electrically or pneumatically disarm, or otherwise secure the valve. PNPS interpretation and practice had been that a "Danger" tag on the control switch meets this requirement for a fail-closed valve with no automatic valve opening signal. At the time, Primary Containment Isolation Valves AO- 5042A and AO-5044A were inoperable in the isolated position but were not deactivated.

The apparent cause of this event is failure to properly revise procedures and practices when LCO TS 3.7.A.2.b was revised in 1988 to permit periodic opening of the deactivated valve in the line under administrative controls.

This event was not risk significant and had no adverse impact on the health and safety of the public.

05000293/LER-2013-00231 January 2014Pilgrim

On Sunday January 20, 2013, at 2050 hours with the reactor at 100% core thermal power (RMSS in RUN), PNPS declared SRV-3B inoperable and entered Technical Specification (TS) 3.6.D.2 requiring an orderly re- actor shutdown such that reactor coolant pressure is less than 104 psig within 24 hours. On Monday January 21, 2013, at 1300 hours (16 hrs and 10 minutes) reactor coolant pressure was lowered to less than 104 psig.

SRV-3B had been declared inoperable consistent with PNPS procedures that state an SRV is inoperable if the first stage pilot thermocouple temperature is 35° F below its baseline temperature. This LER Supplement provides the determination of cause for the leakage. The cause of the SRV leakage was that the natural fre- quency of the pilot assembly was close to a resonant frequency of the valve assembly when installed on the PNPS main steam line, that had failed to be considered in the design of the SRV. A contributing cause was wear and looseness of parts in the main stage of RV-203-3B.

The reactor was depressurized and a new pilot valve assembly was installed on SRV-3B. On January 22, 2013, at 1015 hours reactor restart was commenced. On January 24, 2013 at 0312 hours 100% core thermal power was achieved.

This LER also reports the as-found setpoint of one SRV pilot valve tested was less than the minimum pres- sure required by TS 3.6.D.1.

This event had no impact on the health and/or safety of the public.

05000293/LER-2013-01018 December 2013Pilgrim

On Saturday, October 19, 2013 at 0330 (EDT) with the reactor critical at 1% core thermal power (CTP), and the mode switch in START UP, a high reactor water level condition (+55") resulted in a valid Group I primary containment isolation signal and resultant closure of the main steam isolation valves (MSIV), MSIV drain line valves, and reactor recirculation system sample line valves. The plant was in the process of starting up with reactor pressure at approximately 290 psig with a corresponding reactor coolant temperature for that pressure. Reactor startup was suspended and control rods were manually reinserted. Reactor water level was recovered and maintained within normal bands. Plant systems responded as designed to the Group I containment isolation signal.

The direct cause of MSIV isolation was automatic actuation of the Group I containment isolation signal due to high reactor high water level with the mode switch in START UP. The cause of the high reactor water level was due to unexpected rapid opening of three turbine steam bypass valves. The reason the bypass valves rapidly opened was due to a malfunction of the mechanical pressure regulator (MPR). The MPR malfunctioned because an increased error signal between turbine steam pressure and the MPR setpoint due to friction between the MPR pilot valve and the pilot valve bushing resulting from lack of rotation of the pilot valve bushing. Corrective action was taken to flush the needle valve that controls oil flow to the pilot valve bushing and exercise of the pilot valve bushing to restore proper rotation.

This event posed no threat to public health and safety.

05000293/LER-2013-00913 December 2013Pilgrim

On Monday October 14, 2013 at 2121 hours (EDT), with the reactor critical at 100% core thermal power, the mode switch in RUN, and offsite power 345KV Line 342 out of service for a scheduled upgrade, a loss of offsite power (LOOP) occurred due to the loss of the second 345KV Line 355. All control rods fully inserted, main steam isolation valves closed on the loss of power to the reactor protection system, and the emergency diesel generators automatically started supplying power to both 4160V safety buses. Following the scram, reactor water level lowered to +12 inches initiating the Primary Containment Isolation System (Group II, Reactor Building Isolation System (RBIS); and Group VI - Reactor Water Cleanup System) automatically per design. A plant cool down commenced with reactor water level being maintained in the normal post-scram band of +12 inches to +45 inches utilizing the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems.

The cause of Line 355 loss was due to a failure of an offsite substation tower support. The offsite tower was repaired and Line 355 was energized at 2023 hours on October 15, 2013.

These events posed no threat to public health and safety.

05000293/LER-2013-00821 October 2013Pilgrim

On Thursday, August 22, 2013 at 0755 (EDT) with the reactor critical at 98% core thermal power (CTP), and the mode switch in RUN, the Pilgrim Nuclear Power Station (PNPS) was manually scrammed due to lowering reactor water level resulting from a trip of the reactor feed pumps. The reactor feed pumps tripped due to a loss of power to the pump seal cooling water flow switch relays and resultant automatic actuation of the feed pump trip circuit.

The direct cause of the reactor feed pump trip was an automatic actuation of the feed pump trip circuitry. The feed pump trip circuits actuated as designed in response to a loss of power to the pump seal cooling water flow switch relays. The seal cooling water flow switch relays lost power as a result of a 120V AC breaker trip resulting from a short-to-ground fault in the associated circuits fed by the breaker. Corrective action was taken to repair the ground fault in the associated circuits fed by the 120V AC breaker and to revise reactor feed pump trip circuit design to remove the loss of seal water trip function.

This event posed no threat to public health and safety.

05000293/LER-2013-00716 September 2013Pilgrim

On Tuesday, July 16, 2013 at 1652 (EDT) and again on Wednesday, July 17, 2013 at 1054 (EDT) with the reactor at 100% core thermal power (CTP) the Pilgrim Nuclear Power Station (PNPS) declared the ultimate heat sink (UHS) and the salt service water (SSW) system inoperable due to high sea water inlet temperatures greater than 75°F. A maximum sea water inlet temperature reading of 75.5°F was observed and the maximum duration for either event was 5.5 hours.

The limiting condition for operation (LCO) action for technical specification (TS) 3.5.B.4 was entered then exited based on the rise and fall of sea water inlet temperature. Plant systems and components operated as required and no equipment failures occurred. The plant was not shutdown due to the short duration of the sea water temperature excursion.

The cause of the high sea water inlet temperature readings was sustained increased sea water surface temperature in Cape Cod Bay due to hot summer weather conditions and the contribution from recirculation of water from the plant's outfall due to wind and tidal conditions. Corrective action was completed to establish an operational decision making issue (ODMI) action plan to reduce station power levels prior to reaching the TS UHS LCO temperature limit.

This event posed no threat to public health and safety.

05000293/LER-2013-00418 June 2013Pilgrim

Switch (RMSS) in "Startup/Hot Standby", the turbine generator previously removed from service, and the reactor sub- critical on Intermediate Range Monitors Range 2 and lowering, a manual reactor scram was inserted due to reactor pressure decreasing faster than normal. At the time of the manual reactor scram PNPS was conducting a planned reactor shutdown to commence refueling outage (RFO) -19. All control rods fully inserted and Primary Containment Isolation System (PCIS) Group II (Reactor Building) and Group VI (Reactor Water Cleanup System) actuations occurred as designed due to the expected reactor water level shrink associated with the scram signal. All plant systems responded as designed. Off-site power was unaffected and was supplied by the start-up transformer (normal power supply for refuel and reactor shutdown operations).

The Main Steam Isolation Valves (MSIV) were manually closed to terminate the pressure reduction and the High Pressure Coolant Injection (HPCI) system was manually started in the pressure control mode. The plant cooldown continued with the HPCI system in pressure control and reactor water level maintained within normal bands with the condensate and feedwater system.

The Root Cause of the event was that procedure PNPS 2.1.5 did not limit operation of MO-S-2, Steam Seal Bypass Valve, to below the steam line pressure design operating limit (250 psig) of the steam seal bypass. The procedure was revised to preclude recurrence.

05000293/LER-2013-0038 April 2013Pilgrim

On Friday February 8, 2013, at 2117 hours with the reactor initially at 85% core thermal power, Pilgrim Nuclear Power Station (PNPS) experienced a loss of off-site power (LOOP) resulting in a load reject and a reactor scram. All rods fully inserted and the Emergency Diesel Generators automatically started and powered safety-related buses A5 and A6. All other safety systems functioned as required.

The plant stabilized in Hot Shutdown. At the time of the event a significant winter storm (Nemo) was buffeting Southern New England. At 2200 hours PNPS in conjunction with the local grid operator determined off-site power sources were not reliable and efforts to restore off-site power were temporarily suspended. At 2200 hours, PNPS declared a Notification of Unusual Event. On February 10, at 1055 hours, one of two off-site power supplies was restored, all safety buses were powered from the startup transformer and the Unusual Event was exited. Later on February 10, at 1402 hours with the plant in Cold Shutdown, ice bridging on a startup transformer insulator caused its 345 KV supply breaker to open resulting in a second LOOP. Again the EDG's started and powered safety-related buses. All other safety systems functioned as required. Shutdown cooling was restored at 1426 hours.

On February 10, at 2020 hours, this occurrence was reported to the USNRC as documented in EN# 48739.

The severe winter storm which caused extensive generalized geographical damage to the electrical distribution network was root cause of the LOOP events.

These events posed no threat to public health and safety.

05000293/LER-2013-0017 March 2013Pilgrim

On Thursday, January 10, 2013 at 1534 hour (EST), with the reactor at 100% core thermal power, both reactor recirculation pumps unexpectedly tripped and a manual reactor scram was inserted as required by station procedures. Following the reactor scram, all rods were verified to be fully inserted and the Primary Containment Isolation System Group II (Reactor Building) and Group VI (Reactor Water Cleanup System) actuations occurred as designed due to the expected reactor water level shrink associated with the scram signal. All other plant systems responded as designed. The scram was uncomplicated and decay heat was released to the main condenser via the turbine by-pass valves.

The cause of the two reactor recirculation pumps tripping was due to the inadvertent seal-in of a relay (pump trip interlock) in the Low Pressure Coolant Injection (LPCI) Loop Select Logic circuitry within the Residual Heat Removal (RHR) System during surveillance testing. When the logic was reset at completion of testing, a normally open relay contact (which was inadvertently closed) interlocked with the recirculation pumps circuit, sent a trip signal to their drive motor breakers.

Corrective action has been taken to revise the subject surveillance procedure with steps to reinstall relay covers and added a verifier to observe relay status/ state prior to resetting the relay logic circuit.

This event had no impact on the health and/or safety of the public.

05000293/LER-2011-00721 December 2012Pilgrim

On Monday, December 26, 2011, at 1250 hours, with the reactor at 100% core thermal power, the station entered a 24-hour action statement to initiate a controlled shutdown and be less than 104 psig reactor pressure due to suspected leakage across the first stage of safety relief valve (SRV) RV-203-3D. The SRV was declared inoperable due to criteria specified in a Pilgrim plant procedure. Specifically, the SRV is inoperable if the pilot stage thermocouple temperature is 35° F below its baseline temperature. The safety relief valve was declared inoperable and the Limiting Condition for Operation (LCO) for Technical Specification (TS) 3.6.D was entered. Due to the valve being declared inoperable, the station was required to be shutdown and reactor coolant pressure be below 104 psig within 24 hours per TS 3.6.D.2.

Following the shutdown, RV-203-3D was repaired with a new pilot valve and the plant was returned to full power operation.

This event had no impact on the health and/or safety of the public.

05000293/LER-2011-00630 January 2012Pilgrim

On November 30, 2011, at 1747 hours, with the reactor at 100% core thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNPS) declared the High Pressure Coolant Injection (HPCI) system inoperable due to the HPCI turbine governor control valve (HO-2301-24) failing to open during planned post-maintenance testing. The governor control valve is a hydraulically operated valve and its normal position is closed. The valve has a safety function to open on a demand signal during certain event mitigation scenarios requiring the HPCI system operation.

The governor control valve failed to open during the post-maintenance testing from the Alternate Shutdown Panel (ASP) and subsequently from the Main Control Room. The HPCI system was declared inoperable and the Limiting Condition for Operation (LCO) for Technical Specification (TS) 3.5.C.2 was entered.

This event had no impact on the health_ and/ or safety of the public.

05000293/LER-2011-0038 July 2011Pilgrim

On May 10, 2011, during STARTUP Mode with reactor thermal power at approximately 1.7 percent, Pilgrim Nuclear Power Station (PNPS) experienced an automatic reactor scram while raising reactor temperature and pressure. The reactor scrammed on intermediate range monitor (IRM) hi-hi flux on both reactor protection system channels. Prior to the scram, operators took the reactor critical, reached the point of adding heat, and established a heatup rate. During the heatup, operators observed a high heatup rate and in response, the shift manager directed operators to insert control rods to reduce the heatup rate. The number of rods or number of notches to insert was not specified. Power began to lower as expected; however, operators did not recognize that inserting control rods to reduce heatup rate with rising moderator temperature caused the reactor to become subcritical. After achieving a temperature change from the power reduction, operators withdrew the same control rods before evaluating the core condition. The resultant reactor response was a faster power ascension rate than expected, which led to an automatic intermediate-range high-flux reactor trip. All systems operated as expected, in accordance with design.

The root cause of this event was determined to be the failure to adhere to established standards and expectations due to a lack of consistent supervisory and management enforcement. Investigation into the event revealed several examples of inconsistent enforcement of administrative procedure requirements and management expectations for command and control, roles and responsibilities, reactivity manipulations, clear communications, proper briefings, and proper turnovers.

Numerous corrective actions have been taken or are planned to reinforce standards and retrain crew members on the fundamentals of criticality and heating range operation.

This event had no impact on the health and/ or safety of the public.

05000293/LER-2011-00120 April 2011Pilgrim

At 0055 hours on Sunday, February 20, 2011, the Pilgrim Nuclear Power Station (PNPS) commenced a controlled shutdown of the reactor due to the 'B' train of Reactor Building Closed Cooling Water (RBCCW) being declared inoperable and expected to exceed its 72-hour Limiting Condition for Operability (LCO) as required by TS prior to return to operable status.

With the plant operating at 100% power, leakage of Salt Service Water (SSW) was detected in the RBCCW system due to high chloride levels and increased inventory in the system. An investigation into the event determined that the source of the SSW was isolated to the 'B' RBCCW heat exchanger which is designed to cool RBCCW under normal and post-accident conditions. The quantity of the leakage was determined to exceed the design limits established to ensure post-accident operation of the system and the 'B' train of RBCCW was subsequently declared inoperable.

The leak detection and repair activities identified a single tube leak resulting from an improperly modified tube sleeve (shortened and incorrect bevel) which accelerated wear on the parent tube.

This event had no impact on the health and/or safety of the public.

05000293/LER-2011-00220 April 2011Pilgrim

On Sunday, February 20, 2011 at 1034 EST, with the reactor shutdown and all control rods fully inserted a valid Reactor Protection System (RPS) low reactor water level initiation signal (+12 inches) was received. The RPS actuation signal resulted in a reactor scram and actuation of Primary Containment Isolation System (PCIS) Group II (Drywell) isolation, Group VI (RWCU) isolation and a Reactor Building Isolation System (RBIS) actuation. At the time of the event, a controlled reactor shutdown and cooldown was in progress. The Reactor Mode Selector Switch was in "Startup" and the low reactor water level actuation signal was the result of reactor water level control difficulties during the cool-down using the Mechanical Pressure Regulator (MPR). Reactor water level was immediately restored, the isolations (Group II and VI) were reset; and the RPS signal was reset at 1135 EST. All systems operated as expected, in accordance with design.

Corrective actions taken included the revision of the reactor heat-up / cool-down procedure to incorporate lessons learned and to identify the Bypass Valve Opening Jack (BVOJ) as the preferred method for executing a reactor pressure vessel cool-down. Corrective actions planned include the performing of an analysis of MPR/RPV and level response during plant cool-down at the plant simulator and evaluate results for disposition. This event had no impact on the health and/ or safety of the public.

05000293/LER-2010-00111 March 2010Pilgrim

On January 10, 2010 during a backwash evolution on the Reactor Building Closed Cooling Water (RBCCW) System heat exchanger, plant operators discovered a broken bolt on the pipe clamp for the seismic support for the instrument line of the local pump suction pressure gauge attached to the RBCCW "A" Train pump suction pipe. The seismic support is provided to ensure instrument line integrity and is relied on to ensure that RBCCW leakage limits will not be exceeded. The broken bolt compromised the design function of the seismic , support and RBCCW "A" Train was conservatively declared inoperable until a new bolt was installed. Pilgrim Station was operating at 100% power when the condition Was identified.

Subsequent engineering reviews could not determine the exact time that the bolt broke. Based on the condition of the bolt it was assumed that the bolt was broken for time period that exceeded the 72 hour .

allowable Technical Specification (TS) Limiting Condition of Operation (LCO) action statement for one RBCCW subsystem inoperable.

An immediate corrective action was completed to install a new bolt on the seismic support clamp. Additional corrective actions were taken to identify extent of condition and to walkdown the Reactor Building Auxiliary Bay areas where similar conditions could exist. No other broken bolts or seismic support damage was identified.

The probable apparent cause was identified to be corrosion that caused a progressive crack which eventually failed the bolt.

The event posed no threat to public health and safety.

05000293/LER-2009-00219 February 2010Pilgrim

On December 22, 2009, at 0845 hours, with the reactor at 100% core thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNPS) declared Secondary Containment inoperable due to the loss of a water seal on two 14 inch drain lines in the Torus Compartment designed to mitigate the consequences of a flood in the Reactor Auxiliary Bay. There are two troughs that are designed to provide a floodable volume (the torus room) in the event of a significant water leak in the auxiliary bay, while maintaining secondary containment integrity by ensuring a given water level is maintained in the troughs. The initial assessment of the condition indicated that the cross-sectional area of pipes, as found, have the potential to exceed the analytical value of allowable Secondary Containment leakage pathway size documented in the design calculations.

The Limiting Condition for Operation (LCO) for Technical Specification (TS) 3.7.C.2.a, was "immediately entered at 0845 until the condition was corrected. Immediate corrective actions taken were to refill the trough to the correct level, initiate repairs to the level,Switch, which was found to be defective, and to enhance the weekly tour requirement of the Torus Compartment performed by plant operations. The TS LCO was subsequently exited at 0945 hours, one hour into the allowed 4 hour time frame, at which time Secondary Containment was declared operable.

The event posed no threat to public health and safety.

05000293/LER-2008-00712 February 2009Pilgrim

On Saturday, December 20, 2008 at approximately 1045 hours and while 'n a Hot Shutdown condition from the previous day's reactor scram (Reference LER 2008-006-00), Pilgrim Station experienced a momentary loss of all 345kv off-site power to the Startup Transformer (SUT) X4. As a result, the following safety system automatic actuations occurred: Reactor Protection System (RPS) actuation (all control rods were previously inserted), start of both Emergency Diesel Generators (EDG) and loading of their respective emergency buses, actuation of Primary Containment Isolation Systems (PCIS) Groups I, II, VI and Reactor Building Ventilation. The High Pressure Coolant Injection (HPCI) System was placed in service for reactor pressure control, and Reactor Core Isolation Cooling (RCIC) System was placed in service for reactor level control. All plant systems functioned as designed and expected.

The direct cause of the momentary loss of all 345kv off-site power was a Phase B to ground fault on the switchyard Line 355 bus section (Bridgewater Station) which caused ACB-102 and ACB-103 breakers to trip. The ACB-103 breaker tripped because it received a remote transfer trip signal from Auburn Street Station owned by the transmission system operator, National Grid (NGRID). The ground fault was cleared by the ACB-102 breaker, and the Bridgewater Station breakers (the ACB-105 breaker was already open from the previous day's reactor scram), however, the ACB-103 breaker should not have tripped. Tripping of ACB-102 and ACB-103 resulted in a loss of the SUT and transferring of the safety busses to the EDGs.

Immediate corrective actions taken included a visual inspection for damage which was completed with satisfactory results and a successful carrier test was performed on the Line 342 to and from Pilgrim Station. Additionally, a ground overcurrent relay was reset at the Auburn Street Station. Corrective actions planned include working with local Transmission and Distribution Companies to review and reset line protection relays based on investigation results.

The event posed no threat to public health and safety.