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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.

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.

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.

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.

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.

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.

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.

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.

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.

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

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.