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 Report dateSiteEvent description
05000423/LER-2017-00120 March 2017MillstoneAt 0835 on January 20, 2017 while operating in MODE 1 at 100% power, a door in the auxiliary building at Millstone Power Station Unit 3 failed to fully close following personnel passage due to the failure of the mechanical door closer mechanism. This door is part of the secondary containment boundary which must be intact for the supplemental leak collection release system to perform its safety function. Operators determined that the condition of the door rendered secondary containment inoperable and resulted in a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material and mitigate the consequences of an accident. The door was repaired and the door completely closed at 1256 on January 20, 2017. This condition is being reported under 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D).
05000336/LER-2016-00220 September 2016Millstone

On August 11, 2016 at 09:59 while operating in MODE 1 at 100% power, operators manually tripped the Millstone Power Station Unit 2 reactor due to degraded condenser vacuum caused by the loss of two out of four circulating water pumps (CWPs) ('A' and 'C' CWPs). The reactor trip was uncomplicated and decay heat removal was via the steam dumps to the condenser. All control rods inserted on the reactor trip. All safety systems functioned as required.

The 'A' and 'C' CWPs stopped due to a loss of normal control power in the associated variable frequency drive (VFD) system. Both backup uninterruptible power supplies (UPS) in the VFDs associated with the 'A' and 'C' CWPs failed to function resulting in the 'A' and 'C' CWPs stopping. This coincided with a grid disturbance that occurred during a storm. The loss of two out of four CWPs caused a reduction in condenser vacuum. The operators manually tripped the reactor due to degrading condenser vacuum.

The associated UPSs were replaced.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B).

05000423/LER-2016-0059 August 2016Millstone

On June 12, 2016, with Millstone Power Station Unit 3 operating in MODE 1 at 100% power, operators identified the third stage of the "A" Reactor Coolant Pump (RCP) seal had failed, which resulted in an unidentified Reactor Coolant System leak greater than the Technical Specification (TS) limit of 1 gallon per minute. Operators initiated a plant shutdown as required by TS 3.4.6.2 ACTION Statement b. During the downpower, steam generator levels were not adequately maintained and the Engineered Safety Features Actuation System generated a Turbine Trip and Feed Water Isolation on Steam Generator Water Level - High-High being exceeded on the 'B' steam generator.

In response, operators manually tripped the reactor at 23:37 (MODE 1, at approximately 20 % power). The auxiliary feedwater system started as designed.

Safety systems functioned as expected. There were no radiological challenges as a result of the event. The plant entered COLD SHUTDOWN on June 14, 2016 at 01:29.

The cause of the unidentified leak was a failed third stage on the "A" RCP seal. The seal was replaced.

The cause of the feedwater transient and resultant manual reactor trip was due to operator performance in controlling steam generator level. This is being addressed in the Corrective Action Program.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B). Additionally, the plant shutdown is being reported in accordance with 10 CFR 50.73 (a)(2)(i)(A) as the completion of any nuclear plant shutdown required by the plant's Technical Specifications.

05000423/LER-2016-00413 July 2016Millstone

On May 15, 2016, with Millstone Power Station Unit 3 (MPS3) operating in MODE 1 at 74% power, the operators observed decreasing hydrogen pressure in the main turbine generator. Upon field investigation it was determined there was an active hydrogen leak from the main generator. The operators manually tripped the reactor and vented the hydrogen from the main generator. The reactor trip was uncomplicated. The auxiliary feedwater (AFW) pumps started as designed on low steam generator level and operators maintained steam generator level.

The active hydrogen leak was the direct cause of the manual reactor trip. The hydrogen leak was caused by a dislodged plug on a port on the main generator. MPS maintenance procedures did not contain adequate procedural guidance in that there was no specific direction for installation, i.e., torque value and verifications. The procedures will be revised to include specific direction to tighten the plugs and applicable verifications (i.e., torque value, peer checking). Additional corrective actions are being taken in accordance with the station's corrective action program.

The actuation of the RPS and the automatic start of the AFW pumps is being reported in accordance with 10 CFR 50.73 (a)(2)(iv)(A) as an event that resulted in manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B).

05000423/LER-2016-0038 June 2016Docket Number
Millstone

(MPS3) performed the 'B' train Supplementary Leak Collection and Release System (SLCRS) Negative Pressure Verification Surveillance. This test was completed with unsatisfactory results. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.6.1 was entered. Later, on April 9, 2016, with the reactor in MODE 1 and approximately 24 percent power, MPS3 operators conducted the 'A' train SLCRS Negative Pressure Verification Surveillance with unsatisfactory results. Because both trains of SLCRS failed the required surveillances, TS LCO 3.6.6.2 was entered. Operations continued with the planned plant shutdown associated with the RFO and entered COLD SHUTDOWN, MODE 5 at 1241 on April 10, 2016.

SLCRS did not meet the acceptance criteria due to the aggregate impact of a number of dampers not providing effective isolation.

SLCRS was restored to operable condition prior to entering MODE 4 when starting back up from RF017. Detailed procedural steps are being developed for verifying the closed position of SLCRS isolation dampers and being incorporated into the damper post maintenance testing matrix. Associated surveillance procedures are being revised to improve monitoring program for SLCRS isolation dampers.

Since both trains of SLCRS failed to meet TS acceptance criteria, this condition is reportable pursuant to 10CFR50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of a safety function for systems or structures to control the release of radioactive material, and 10CFR50.73(a)(2)(v)(D) to mitioate the consequences of an accident.

05000423/LER-2016-00223 March 2016Millstone

On January 25, 2016 an automatic reactor trip occurred on Millstone Power Station Unit 3 (LER 2016-001-00). Following the Millstone Power Station Unit 3 reactor trip, (Mode 3, 0% power), while operators were performing feedwater isolation actuation verification steps, it was identified that the 'C' Feedwater Isolation Valve, 3FWS*CTV41C, did not auto-close as expected. Operators subsequently closed 3FWS*CTV41C from the control room.

Troubleshooting identified that a temporary jumper, associated with a maintenance activity completed in November 2014, was inadvertently left installed due to a human performance error. This jumper bypassed the input from the Solid State Protection System to isolate 3FWS*CTV41C upon receipt of an actuation signal. This input is required by Technical Specification 3.3.2 and Table 3.3-3, Engineered Safety Features Actuation System Instrumentation. The function to automatically isolate 3FWS*CTV41C was thus defeated. However, the safety function to isolate feedwater was met due to redundant valves in series that operated correctly via the 'C' feed regulating valve and the 'C' feed regulating bypass valves. Also, the containment isolation function was not lost as the valve, 3FWS*CTV41C, was capable of remote operation from the control room.

This condition is reportable in accordance with 10 CFR 50.73 (a)(2)(i)(B) as any operation or condition that is prohibited by the plant's Technical Specifications. The jumper was removed and the valve was verified to be OPERABLE.

05000423/LER-2016-00116 March 2016Millstone

On January 25, 2016 at 0147 hours an automatic reactor trip occurred at Millstone Power Station Unit 3 while the unit was in Mode 1, operating at 100 percent power due to a trip of the 'B' Reactor Coolant Pump. The 'B' Reactor Coolant Pump tripped on a ground fault, which in turn caused the Unit 3 reactor to trip on reactor coolant system low loop flow (Reactor Protection System actuation). All control rods fully inserted into the reactor. The auxiliary feedwater pumps started as designed on low steam generator level and operators maintained steam generator level.

All other post trip actions were standard and all safety systems operated as expected.

The 'B' Reactor Coolant Pump trip was caused by failure of one of three motor capacitors on the Reactor Coolant Pump motor. The failed capacitor on the affected Reactor Coolant Pump motor was replaced and the integrity of the motor windings, cabling, breaker, and associated protective relays were verified. Based on an assessment of the event, the risk impact was determined to be very small. There were no radiological challenges and the health and safety of the public were not affected.

The actuation of the reactor protection system and the automatic start of the auxiliary feedwater pumps is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B).

05000336/LER-2015-0037 January 2016Millstone

On November 8, 2015 during reactor start-up from refueling outage 2R23 with the reactor in mode 1 at approximately 57.5 percent power the operators received indication of an oil leak on the 'C' reactor coolant pump (RCP) motor lower oil reservoir. The operators monitored the 'C' RCP oil level and bearing temperatures. Upon noting the RCP oil level dropping at a rate of approximately 1.7percent per hour and the lower RCP guide bearing temperature rising, the operators entered abnormal operating procedure (AOP) 2575, Rapid Downpower, and commenced a rapid down- power. The operators manually tripped the reactor at approximately 19percent power in accordance with AOP 2575 and entered emergency operating procedures. The reactor trip was uncomplicated. All safety systems operated as designed.

The cause of the valid actuation of the reactor protection system (RPS) was the operators manually tripping the reactor via the manual reactor trip push button switches. The AOP was entered due to an oil leak from the tubing to the level transmitter on the 'C' RCP motor lower oil reservoir. The tubing failure mechanism was determined to be high cyclic fatigue. Repairs were completed, the unit was restarted, and returned to service. Additional corrective actions are being taken in accordance with the station's corrective action program.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A), as any event or condition that resulted in manual or automatic actuation of any of the systems (RPS) listed in paragraph 10 CFR 50.73 (a)(2)(iv)(B).

05000336/LER-2015-00211 December 2015Millstone

On June 11, 2015, while Millstone Power Station Unit 2 (MPS2) was in MODE 1 operating at 100 percent power, Engineering identified that due to a degraded check valve, the post-accident radioactivity release rates assumed in the FSAR could be affected. While performing 'B' High Pressure Safety Injection (HPSI) pump in-service testing, the measured flow was lower than expected. Because both trains of HPSI, Low Pressure Safety Injection (LPSI) pumps, and Containment Spray (CS) pumps share a common minimum flow recirculation line back to the Refueling Water Storage Tank, back-leakage through one of the idle pump's recirculation check valves was postulated as the cause of the observed drop in recirculation flow. Troubleshooting was performed, and it was determined that the minimum flow check valve associated with the 'A' CS pump, 2-CS-6A, was back-leaking. The associated minimum flow isolation valve was closed to eliminate the flow path. Back-leakage of the minimum flow recirculation check valves on the HPSI, LPSI, and CS pumps was not previously considered in radiological release analysis. This leakage had the potential to adversely affect calculated post-Loss of Coolant Accident recirculation phase radioactivity release rates under some postulated scenarios.

This event is being reported as an event or condition that could have prevented fulfillment of a safety function to control the release of radioactive material under 10 CFR 50.73(a)(2)(v)(C). The cause of the event was a failed open check valve. As a corrective action, the failed check valve 2-CS-6A has been repaired.

NRC FORM 356 (02-2014) RC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (02-2o14) APPROVED BY OMB: NO. 3160-0104 EXPIRES: 0113112017 Reported lessons learned are Incorporated into the licensing process and fed back lo industry.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to Intocollects.Resourcei10 CFR nrc.gov, and to the Desk Officer, Office of Information end Regulatory Affairs, NEOB-t0202, (3/5041104), Office of Managemen1 and Budget, Washington, DC 20503.11 a means used to impose an information collection does not of-splay a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not requi red to respond lo, the Information collection.

1. EVENT DE3CRIIPT

  • On June 11, 2015, while Millstone Power Station Unit 2 (MPS2) was in MODE 1 operating at 100 percent power, Engineering identified that due to a degraded check valve, the post-accident radioactivity release rates assumed in the FSAR could be affected. While performing 'B' High Pressure Safety Injection (HPSI) pump in-service testing, the measured flow was lower than expected. Because both trains of HPSI, Low Pressure Safety injection (ILPSO pumps, and Containment Spray (CS) pumps share a common minimum flow recirculation line back to the Refueling Water Storage Tank (RWST), back-leakage through one of the idle pumps recirculation check valves was postulated as the cause of the observed drop in flow. Troubleshooting was performed, and it was determined that a degraded minimum flow check valve associated with the 'A' CS pump (2-CS-6A) was back-leaking into the “A" train suction line and flowing to the RWST through 2-CS-14A (RWST suction check valve) and normally open 2-CS-13.1A (RWST suction line isolation).

The associated minimum flow isolation valve for 2-CS-6A was closed to eliminate the path and the valve was repaired.

Subsequently, engineering evaluation determined that during a loss of power to one facility or train of the Emergency Core Cooling System (ECCS), this back-leakage flow path could result in more leakage to the RWST than had been previously considered in the accident analysis. This leakage was evaluated for the potential to adversely affect calculated post—Loss of Coolant Accident recirculation phase radioactivity release rates under some postulated scenarios. Specifically, without power on the affected (i.e.: back-leaking) train, the CS discharge flowpath would not automatically open resulting in leakage to the RWST.

Prior to this discovery, the East time that the 'A' containment spray pump was run was on April 15, 2015, which would have opened 2-CS-6A. On this basis, it was concluded that this condition existed from April 15, 2015 to June 11, 2015. During this period, the 'A' Emergency Diesel Generator (EDG) was inoperable 4 times for a total of approximately 25 hours. However, all 4 times were for surveillances. At no time during this period was any maintenance done on the 'A' EDG. It is judged that, between accident initiation and commencement of sump recirculation (approximately 2 hours with only one train available), the EDG could have been made available.

This event is being reported as an event or condition that could have prevented fulfillment of a safety function to control the release of radioactive material under 10 CFR 50.73(a)(2)(v)(C). Note: The initial non-emergency report (#51149) of this issue on June 11, 2015 was subsequently updated on July 10, 2015. An additional unidentified release path from the original design of the plant was reported to the NRC as a follow-up notification on July 10, 2015.

BACKGROUND

The MPS2 Containment Spray (CS) system functions as an engineered safety feature to limit containment pressure and temperature after a loss-of-coolant accident (LOCA) and Main Steam Line Break (MSLE) accident, and thereby reduce the potential for leakage of airborne radioactivity to the outside environment. A minimum flow recirculation line is included in the design for recirculating water from the outlet of the pump to the RWST.

With the CS system operating during a design basis accident, a small portion of the CS pump discharge flow recirculates to the RWST during the injection phase; however, the recirculation line is isolated from the RWST when transferring to sump recirculation. All seven ECCS/CS pumps (2 LPSI, 2 CS, and 3 HPSI) have minimum flow recirculation lines that tie into one common header to the RWST. Exhibit A attached to this LER is a sketch depicting the configuration.

Evaluation of this failure identified that during a SBLOCA (Small Break Loss of Coolant Accident) concurrent with a loss of power to the "A" train associated with the leaking 2-CS-6A, the operating (opposite) train HPSI pump would pressurize the common recirculation header during the injection phase of the accident as was described above. In the operator response to this postulated event when the 2-CS-6A leaks to the RWST, suction 2-CS-13.1B is manually isolated terminating the release to the RWST. However, it was additionally identified that the "A" train suction header would pressurize because of the continued back leakage with none of the "A" train pumps flowing due to loss of power and no open flowpath (to containment spray or RWST). This pressurization would continue, exceeding the design pressure of the suction header (60 psig) ultimately reaching 500 psig, the lift setpoint for the two "A" train shutdown cooling heat exchanger relief valves downstream of the "A" LPSI and CS pumps. These relief valves discharge to the EDST (Equipment Drains Sump Tank) which is located outside the filtered ventilation boundary. Upon emptying of the RWST and initiation of sump recirculation, the described back-leakage flowpath would contain sump fluid. This additional unidentified release path from the original design of the plant was reported to the NRC as a follow-up notification on July 10, 2015.

For this accident and radiological release sequence to occur the following are required:

  • 2-CS-6A back-leakage
  • SBLOCA
  • Loss of Offsite Power (LOOP)
  • Loss of the "A" train of onsite electric power
  • SBLOCA break size that results in RCS re-pressurization and sustained RCS pressures above 500 psig, such that HPSI would continue to pressurize the recirculation header
  • Fuel damage While the discussion above relates to 2-CS-6A and the "A" train, back leakage from any of the minimum flow check valves and a concurrent loss of power to that train will have a similar outcome.

2. CAUSE:

The cause of the event was a failed-open check valve. Leakage of these check valves was not considered as a failure mode in the Millstone Unit 2 FSAR or original design basis.

3. ASSESSMENT OF SAFETY CONSEQUENCES:

An Operability Determination has been prepared and approved to document the radiological consequences of the condition. The radiological consequences of ECCS/CS min-flow check valve(s) leakage were evaluated for 5 gallons per minute of post LOCA containment sump liquid relieving to the Equipment Drains Sump Tank. This evaluation considered a 30 day unfiltered ground release and used a best estimate source term (gap release, 1% partitioning, etc.). This leakage was added to the consequences of previously identified leak paths, which had been analyzed using design basis LOCA analysis assumptions. The maximum train as found leakage was .089 gpm, well below the 5 gpm limit using best estimate source term.

Even with the addition of the minimum-flow leakage, the control room and offsite dose consequences are within the regulatory limits of 10CFR50.67,

4. CORRECTIVE ACTION:

During the recently completed 2R23 refueling outage, leak testing of ail MPS2 ECCS/CS min-flow check valves was completed. The testing was performed at 500 psig. The results were scaled to 1200 psid to reflect the highest HPSI discharge pressure. The resultant maximum as-found leak rate was 0.089 gpm through the 'A' ECCS/CS train. The failed check valve 2-CS-6A has been repaired. The 'B' HPSI check valve 2-SI-422, the check valve with the highest as-found leakage rate of 0.085 gpm, was repaired and retested. The maximum ECCS/CS pump min-flow as-left check valve leak rate in a train is 0.004 gpm.

Additional corrective actions are being taken in accordance with the station's corrective action program.

5. PREVIOUS OCCURRENCES:

  • None 6. Energy Industry Identification System (EIIS) codes:
  • Pump — P
  • Valve — V
  • Tank —T
  • Containment Spray — CS
  • Safety Injection — Si
05000336/LER-2015-00116 June 2015Millstone

On April 17, 2015, with Millstone Power Station Unit 2 (MPS2) at 100% power and in operating MODE 1, the Nuclear Regulatory Commission questioned the historic operability of Millstone Power Station Unit 2's (MPS2) "C" Reactor Building Closed Cooling Water (RBCCW) pump because several condition reports had reported the need to frequently add oil to the pump's outboard bearing bubbler between November 2014 and early April 2015. Upon further review, Dominion concluded that operability could not be assured since there may not have been sufficient oil to meet mission time requirements at all times with no compensatory measures in place. Plant Technical Specification 3.7.3.1 "Reactor Building Closed Cooling Water System" requires two reactor building closed cooling water loops shall be operable.

Since there were periods of time that plant operators were crediting the "C" RBCCW pump for one of the two required reactor building closed cooling water loops, and operability could not be assured at all times, Dominion is reporting this event pursuant to 10 CFR 50.73(a)(2)(i)(B), as a condition that was prohibited by the plant's Technical Specifications.

The cause of the event was that Administrative procedure OP-AA-102, Operability Determination, Attachment 1 contained guidance that allowed that oil leaks would not require an operability determination under certain conditions.

Procedure OP-AA-102 has been revised to specify that any time a pump's mission time cannot be met due to oil leakage, the pump should be declared inoperable.

05000423/LER-2015-00120 April 2015Millstone

On February 19, 2015, with Millstone Power Station Unit 3 (MPS3) at 100% power and in operating mode 1, an individual on a fire watch rove processed through a dual train high energy line break (HELB) door normally and upon checking the door after passage the individual noted the door did not latch. The Control Room was promptly notified./ An operator was dispatched to investigate. The operator exercised the door lock-set mechanism freeing the latch allowing the door to properly latch. The door was inoperable for approximately 7 minutes. Technical Specification 3.0.3 was entered and exited appropriately.

Although no definite failure mechanism was identified, the door was experiencing high usage due to compensatory fire watch roves entering/exiting the door. The door lockset mechanism was manually manipulated and then tested several times satisfactorily by maintenance personnel. Further, the door design has the door swing such that the HELB event would act to open the door when the lockset mechanism fails. Engineering is evaluating the adequacy of the preventive maintenance frequency. Additionally, a design change to reverse the door swing such that the HELB event would cause the door to close and thus not rely on the lock-set mechanism is being considered. Additional corrective actions are being taken in accordance with the station's corrective action program.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(v)(D), as a condition that could have prevented the fulfillment of a safety function for systems needed to mitigate the consequences of an accident.

05000336/LER-2014-00724 September 2014Millstone

(MPS2) operators commenced an orderly shutdown of MPS2 since turbine driven auxiliary feedwater (TDAFW) pump repairs and required testing would not be completed within the time frame required by Technical Specification Action Statement (TSAS) 3.7.1.2 action c. MPS2 entered operating mode 4 (hot shutdown) at 0738 hours on July 27 thus completing the shutdown as required by the TSAS.

The direct cause of the shutdown of MPS2 was the inability to identify the cause of the surveillance test failure and complete the repair and retesting activities within the allowed TSAS time. Subsequent troubleshooting found foreign material inside the TDAFW recirculation orifice. The foreign material was removed, the TDAFW pump was retested and satisfactorily passed the required TS surveillance test.

Additional inspection did not find any additional foreign material. Additional corrective actions are being taken in accordance with the station's corrective action program.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(A) - The completion of any nuclear plant shutdown required by the plant's technical specifications.

05000336/LER-2014-00624 July 2014Millstone

On May 25, 2014, with Millstone Power Station (MPS) Unit 2 and 3 (MPS2 and MPS3) in operating MODE 1, operating at 100% reactor power, a total loss of offsite power occurred in the MPS switchyard. Both MPS2 and MPS3 experienced a turbine trip on power to load imbalance followed by an automatic reactor trip. Both units' safety related emergency diesel generators (EDGs) automatically started and supplied power to their respective safety busses. As designed, the motor driven auxiliary feedwater (AFW) pumps automatically started on MPS2 and all AFW pumps automatically started on MPS3.

MPS declared an Unusual Event (UE) following the reactor trips and the NRC was notified (EN50142 for MPS2 and EN 50141 for MPS3). Following stabilization of both MPS2 and MPS3, the UE was terminated.

With both MPS2 and MPS3 at full power and one of the Transmission Owner (TO) 345 kV lines out of service for scheduled work, a phase to ground fault occurred on one of the phases on a motor operated disconnect (MOD) on one of the three remaining TO 345 kV lines. The fault was caused by an insulator failure of a MOD switch at a TO substation (offsite). This fault was also sensed as an instantaneous ground in one of the two remaining TO 345 kV lines resulting in this line also tripping. With three of the four TO 345 kV lines out of service, the MPS total electrical output overloaded the single remaining TO 345 kV Line which then also tripped resulting in a total loss of offsite AC power to MPS.

The TO replaced the faulty components and is investigating additional corrective actions. Additional corrective actions are being taken in accordance with the station's corrective action program.

This event is being reported per 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B)(1), (6), and (8). Actuations of the reactor protection system, the AFW system, and the EDGs are reportable under this paragraph.

05000336/LER-2014-00516 July 2014Millstone

At 1933 on May 16, 2014 while in MODE 3, Millstone Power Station Unit 2 (MPS2) exceeded the Limiting Condition for Operation (LCO) of plant Technical Specification (TS) 3.6.2.1 'Containment Spray System' Action a.1 for an inoperable containment spray pump. The 'A' containment spray (CS) pump was declared inoperable at 0018 on May 17, 2014, the date of discovery, following completion of surveillance testing to determine the presence of gas voids.

However, the gas was introduced earlier during the refueling outage and the TS LCO went into effect upon first entry into MODE 3 greater than 1750 psia on May 13, 2014, at 1933. TS 3.6.2.1 Action a.1 requires that the pump be restored to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1750 psia within the following 6 hours. MPS2 had been in a MODE where the CS system was required to be OPERABLE for 70.5 hours prior to completion of the testing. The gases were successfully removed by venting and the system was restored to OPERABLE status at 1221 on May 17, 2014.

This condition is being reported as any operation or condition which was prohibited by the plant's technical specifications in accordance with 10 CFR 50.73 (a)(2)(i)(B). The condition was caused by not adequately venting the CS system and delays in communicating the surveillance results, combined with a need to schedule performance of the surveillance testing earlier in a refueling outage. Corrective actions planned will improve scheduling of the testing and will result in more timely communications of the results from completed testing.

05000336/LER-2014-0049 June 2014Millstone

On April 10, 2014, with Millstone Power Station Unit 2 in MODE 6 at 0% reactor power, while de-terminating the motor leads for the 'B' Motor Driven Auxiliary Feedwater (MDAFW) Pump motor, foreign material (ty-wraps and a plastic bag) were found inside the cable environmental seal (Raychem boot) of the 'A' phase. An inspection of the electrical motor leads revealed no damage occurred. This motor was last re-terminated In May 2000. The motor leads to the other phases did not have any foreign material inside the Raychem boot. Since the Raychem boot was not in the as tested environmentally qualified (EQ) configuration the 'B' MDAFW pump was considered inoperable.

Plant Technical Specifications (TS) 3.7.2.1 Action d, requires, if two AFW pumps are inoperable in operating MODES 1, 2, and 3, the plant must be placed in at least HOT STANDBY within six hours and in HOT SHUTDOWN within the following 12 hours.

A review of the control room logs for the past three years determined there were 4 occasions where there were two AFW pumps inoperable for longer than allowed by TS.

The direct cause was an historical inappropriate maintenance practice which rendered the MDAFW pump inoperable.

The 'A' phase motor lead was subsequently properly re-terminated.

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

05000336/LER-2014-0032 June 2014Millstone

On April 5, 2014, with Millstone Power Station Unit 2 at 100% power in Mode 1, the Enclosure Building was rendered inoperable during a maintenance activity. The condition was created at approximately 1025 and restored to operable status at 1300 on April 5, 2014. Plant Technical Specification (TS) 3.6.5.2 requires that the Enclosure Building shall be operable in Modes 1, 2, 3, and 4. The TS 3.6.5.2 Action is to restore the Enclosure Building to operable status within 24 hours or be in cold shutdown within the next 36 hours.

Therefore, the TS 3.6.5.2 Action requirements were met.

Since the Enclosure Building was inoperable from approximately 1025 until restored at 1300, the safety function of the Enclosure Building to limit radiological releases or mitigate the consequences in the event of a design basis accident could not be assured. The condition was created as the result of a human performance error while performing a maintenance activity associated with the Enclosure Building boundary.

Since this was a human performance error, the event and lessons learned were communicated to the Millstone Site as a Station Human Performance Clock Reset.

05000423/LER-2014-0025 May 2014Millstone

While Millstone Power Station Unit 3 (MPS3) was operating at 100 percent power in Mode 1 on March 12, 2014, operations determined that a concern identified by Engineering as a result of recent industry operating experience (NRC Event Number 49889) could result in an unanalyzed condition. Engineering's review of the direct current (DC) circuits associated with the main turbine emergency lube oil and main generator emergency seal oil pumps determined the described condition to be applicable to MPS3.

This condition is reportable pursuant to 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety. A prompt notification per EN49903 was submitted to the NRC on March 12, 2014.

The cause of the condition was a manufacturer's design error that was installed during plant construction.

Compensatory fire watches have been implemented for affected areas of the plant. Plant modifications are planned that will protect or isolate the affected circuits leaving the initial fire area.

05000336/LER-2014-00131 March 2014Millstone

On January 30, 2014 at 2313, with Millstone Power Station Unit 2 (MPS2) in MODE 1 and at 100% reactor power, the circuit breaker for the group 1 pressurizer proportional heaters tripped. With the 'B' emergency diesel generator out of service for its maintenance outage and therefore inoperable, Technical Specification Action Statement (TSAS) 3.4.4 action b was entered. TSAS 3.4.4 action b requires MPS2 be in at least HOT STANDBY (MODE 3) with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN (MODE 4) within the following 6 hours, unless one group of proportional heaters is restored to operable status. MPS2 completed the shutdown to MODE 3 at 0457 and restored the group 1 proportional heaters to operable status at 0747 on January 31, 2014. The initiation of the shutdown was reported to the NRC (event number 49779) pursuant to 10 CFR 50.72(b)(2)(i) - The initiation of any nuclear plant shutdown required by the plant's technical specifications.

After troubleshooting, faulty heater leads were lifted to remove a failed heater from service. Following appropriate testing, the group 1 proportional heater and the pressurizer were declared operable and MPS2 was subsequently returned to service.

This condition is being reported pursuant to 10 CFR 50.73(a)(2)(i)(A) - The completion of any nuclear plant shutdown required by the plant's technical specifications.

05000336/LER-2013-00419 December 2013Millstone

On November 9, 2013, at 1514, Millstone Power Station Unit 2 (MPS2) experienced a turbine trip and an automatic reactor trip from 95% power MODE 1.due to loss of condenser vacuum. The Unit was in the process of condenser backwashing operations when condenser vacuum was lost due to unexpected pump ramp-down of the 'C' circulating water pump (CWP) when the 'D' CWP was secured as required by procedure.

All control rods inserted on the reactor trip. An auxiliary feedwater (AFW) automatic start occurred post trip, as expected per design, and all other safety systems functioned as required.

The direct cause of the event was the MPS2 'C' CWP ramped off due to failure of contacts on a time- delay relay to deenergize as designed.

The defective relay was replaced. Additional corrective actions are being taken in accordance with the station's corrective action program.

This event is being reported per 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B). Actuations of the reactor protection system and the AFW system are reportable under this paragraph.

05000336/LER-2012-00319 December 2013Millstone

On October 15, 2012 during a Millstone Power Station Unit 2 beyond design basis flooding walkdown, with the unit shutdown in MODE 5 (cold shutdown), a total of 20 four (4) inch and 2 two (2) inch unsealed electrical conduits, con- necting the service water pump room of the intake structure and the turbine building, were identified. These conduits penetrated the wall above the design basis flood height of 22 feet mean sea level (MSL), but below the maximum standing wave height of 26.5 feet MSL inside the intake structure service water pump room. Because these conduits were unsealed at both ends, this condition could have resulted in flooding of the turbine building such that it could have rendered the turbine-driven auxiliary feedwater (AFW) pump inoperable. Additionally, several other unsealed conduit penetrations, conduit penetrations with damaged seals, or wall cracks were identified within the design basis flood zone during these walkdowns. Engineering analysis concluded these additional leak paths would have no impact to equipment needed to perform during the design basis flood without the concurrent intake structure standing wave.

However, there was a potential to affect the functionality of the AFW pumps, the power operated relief valves and the high pressure injection system if the standing wave condition occurred, as assumed, for a duration of one hour at 26.5 ft.

MSL concurrent with the design basis maximum storm surge. Engineering concluded this event was of very low safety significance. These deficiencies were historical in nature and appear to be original construction deficiencies. Upon discovery the identified deficiencies were repaired to restore the design basis for flood protection.

This condition is being reported under 10 CFR 50.73(a)(2)(v)(A) and (B). Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to shutdown the reactor and maintain it in a safe shutdown condition, or remove decay heat.

05000336/LER-2012-00119 December 2013Millstone

At 19:30 on June 7, 2012, with Millstone Power Station Unit 2 (MPS2) operating at 100 percent power in Mode 1, it was determined that a series of gaps in a high energy line break (HELB) barrier rendered equipment in the west 480VAC switchgear room inoperable. The Shift Manager entered Technical Specification (TS) 3.8.2.1 Action, TS " and TS 3.3.3.5 Action "a". The openings were sealed and the equipment restored to operable 3.8.2.1A Action "c status at 16:05 on June 8, 2012. At the time of discovery, appropriate and timely actions were taken that met the TS action requirements. There is evidence that this condition has existed since initial construction. Since the HELB barrier required to maintain operability of bus 22E, Inverters 5 and 6, and remote shutdown panel C-21 was non-functional for an extended period of time, the Actions for TS 3.8.2.1, TS 3.8.2.1A and TS 3.3.3.5 were historically not met. This condition is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B) any operation or condition prohibited by the plants technical specifications" The apparent cause, dating back to original construction, is that no construction details were developed for sealing Q-decking when the ribbing is perpendicular to the top of the wall. As noted above, the gaps were subsequently sealed and an extent of condition inspection was conducted. Additional gaps in the same wall did not affect safety related equipment. No other gaps were identified. This report assumes the steam from a HELB in the turbine building would have affected the operability of safety related equipment in the west 480VAC switchgear room.

Upon further engineering analysis, it was determined that for limited exposure times safety functions could have been prevented for certain postulated high energy line breaks. Therefore, this condition is also being reported pursuant to 10 CFR 50.73(a)(2)(v)(A),(D). I

05000336/LER-2013-00318 December 2013Millstone

While Millstone Power Station Unit 2 (MPS2) was operating at 100 percent power in Mode 1 on October 30, 2013, operations determined that a concern identified by Engineering as a result of recent industry operating experience could result in an unanalyzed condition. Engineering's review of the impact of unfused direct current (DC) ammeter circuits in the Control Room determined the described condition to be applicable to MPS2. It is postulated that a fire could cause one of the DC ammeter wires to hot short to ground. Concurrently, the fire causes another DC wire from the opposite polarity on the same battery to also short to ground. This would cause a current path through the unfused ammeter cable. The original plant wiring design and associated circuitry analysis for the DC ampere indications do not include overcurrent protection features to limit the fault current in this scenario.

This condition is reportable pursuant to 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety. A prompt notification per EN49487 was submitted to the NRC on October 30, 2013.

The cause of the condition was a latent design error that was made during plant construction.

Compensatory measures, i.e., fire watches, have been implemented for affected areas of the plant.

Plant modifications are planned that will protect or isolate the affected circuits leaving the initial fire area.

05000336/LER-2013-0026 May 2013Millstone

On March 7, 2013, with Millstone Power Station Unit 2 in MODE 1 at 100% power, during a turbine driven auxiliary feedwater (TDAFW) pump surveillance test, the operators left the door to the TDAFW pump room open. This condition existed for approximately 30 minutes. The door is a high energy line break (HELB) barrier. With the door open there is no HELB protection for the motor driven auxiliary feedwater (AFW) pumps thus potentially rendering both trains of AFW inoperable. Additionally, during the surveillance run the TDAFW pump was declared inoperable.

The HELB requirements during operation of the TDAFW pump were not fully considered during the pre-job brief.

Management has reinforced expectations to Operations Department personnel and station personnel on the need to ensure the requirements for any barrier are met prior to opening any door or duct. Training for Operations, Maintenance and Engineering is planned to reinforce understanding of expectations concerning passive design features, like flood barriers, fire doors and HELB barriers. Additional corrective actions are being taken in accordance with the station's corrective action program.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(v) as any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat; and (D) Mitigate the consequences of an accident.

05000423/LER-2013-00115 April 2013Millstone

On February 15, 2013, with Millstone Power Station Unit 3 operating in MODE 1 at 100 percent power, operators discovered that the insulating cover installed over 3MSS*PT526, "B" Steam Generator Pressure Transmitter (SGPT) Channel 3, was not properly secured. When found, it was incorrectly assessed as not impacting equipment operability. However, on February 19, 2013, upon further review by the Electrical Equipment Qualification program engineer, it was determined that the SGPT would not be adequately protected from a postulated high energy line break (HELB). Therefore, 3MSS*PT526 was declared inoperable per Technical Specification (TS) 3.3.2, Engineered Safety Features Instrumentation, on February 19, 2013. Information exists that 3MSS*PT526 was inoperable on February 15, 2013, when the problem was initially discovered. Since the condition existed for longer than allowed by TS 3.3.2, Action 20 (i.e., for more than six hours without the bistable being placed in trip), this is a reportable condition per 10 CFR 50.73 (a)(2)(i)(B) as an event or condition which was prohibited by the plant's Technical Specifications.

Lack of awareness of the purpose of the enclosure boundary combined with the lack of clear labeling identifying the HELB enclosure resulted in a delay in declaring the equipment inoperable. Operations created and installed a new label for the insulating box around each SGPT clearly indicating the box is a HELB barrier and is part of the environmental qualification boundary.

05000336/LER-2013-00112 March 2013Millstone

At 0938 on January 15, 2013, with Millstone Power Station Unit 2 (MPS2) in Mode 1, operating at 100 percent power, while performing a monthly required channel check on the safety valve's position indication, technicians discovered that the channel B primary safety valve acoustic valve monitor system (AVMS) position indicator failed to meet acceptance criteria contained in Surveillance Procedure SP2410A. In this instance, the technicians had expected to observe a frequency peak at or near 74.25 Hz on the AVMS' output, which is induced by an operating reactor coolant pump. After reviewing a printout obtained from the instrumentation loop, it was determined that the frequency peak expected at or near 74.25 Hz was not detected. Therefore, the AVMS channel was declared inoperable and Operators entered Technical Specification (TS) 3.3.3.8, Accident Monitoring Instrumentation, Action 3, for Item 6, Safety Valve Position Indicator, which specifies that alternate indications shall be checked once each shift. On January 17, 2013, upon reviewing the results from the previous month's completed channel check surveillance performed in December 2012, it was discovered that the instrument had also failed to meet the acceptance criteria for a channel check in SP2410A at the time. However, it was not recognized that the surveillance results did not meet acceptance criteria, so the instrument was not declared inoperable at the time.

This condition is being reported pursuant to 10 CFR 73(a)(2)(i)(B) as any operation or condition which was prohibited by the plant's Technical Specifications.

The cause was determined to be inadequate procedure guidance in SP2410A which led to confusion when interpreting surveillance results. The surveillance procedure has been revised to include clear acceptance criteria for a channel check.

05000336/LER-2011-00530 January 2012Millstone

At 1230 on December 3, 2011, with Millstone Power Station Unit 2 operating at 100 percent power in Mode 1, the control room was notified that a door sweep became dislodged on a door credited as a boundary door for the Enclosure Building. Operators declared the door inoperable and entered the Action for plant Technical Specification (TS) 3.6.5.2 Enclosure Building at 1235. TS 3.6.5.2 Action requires that the Enclosure Building be restored to operable status with 24 hours or be in Cold Shutdown within the next 36 hours. Repairs to the door were completed and operators exited the Action at 1524 on December 3, 2011. The last known time where the door sweep was not dislodged was within 18 hours of the time of discovery.

The operability of the Enclosure Building ensures that the release of radioactive materials from the primary containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. Since there is no bounding analysis on the impact of this size opening on the ability to complete the safety function, this condition is being reported pursuant to 10 CFR 50.73(a)(2)(v)(C) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

The direct cause of the door seal failure was that the mounting hardware (screws) loosened and fell out. The apparent cause is the door was not being properly maintained. The door was repaired. Doors that are part of the Enclosure Building boundary have been added to the preventive maintenance program.

05000336/LER-2011-0047 November 2011Millstone

On, September 14, 2011, with Millstone Power Station Unit 2 (MPS2) in Model, at 100 % power, a historical condition was discovered where the surveillance procedure used to test the reactor trip circuit breakers (TCBs) specified the TCBs be closed (i.e., applying power to the control rod drive motors thus enabling the control element assemblies (CEAs) to be capable of being withdrawn) prior to demonstrating the TCBs to be OPERABLE. This is contrary to the requirements of Technical Specification (TS) 4.0.4 which prohibits entry into an operational mode or other specified condition unless the surveillance requirement(s) have been performed within the stated interval. This condition is being reported pursuant to 10 CFR 73(a)(2)(i)(B) as an operation or condition prohibited by TS.

The cause was determined to be an inadequate procedure review. During the implementation process for a 2003 TS change, when the surveillance procedure was revised, it was not recognized that the surveillance procedure test method for the TCBs allowed MPS2 to enter a mode or condition prior to demonstrating the TCBs to be operable as required by TS 4.0.4. The surveillance procedure is being revised to specify an alternate method for testing the TCBs that does not result in rendering the CEA drive system capable of CEA withdrawal prior to closing the TCBs for performance of Matrix Logic and Trip Path testing.

05000336/LER-2011-00326 October 2011Millstone

At 09:31 on September 3, 2011, with Millstone Power Station Unit 2 operating at 100 percent power in Mode 1, the "A" train service water loop was declared inoperable when leakage from a degraded service water spool piece degraded beyond pre-established limits. Plant Technical Specification 3.7.4.1 stipulates with one service water loop inoperable, restore the inoperable loop to operable status within 72 hours or be in cold shutdown (Mode 5) within the next 36 hours. Since the leak was unisolable, operators commenced a plant shutdown.

Cold Shutdown Mode 5 was entered at 17:03 on September 4, 2011.

The direct cause of the service water leak was a degraded coating on the piping flange located in the "A" train 10-inch service water line to the emergency diesel generator heat exchangers. The degradation mechanism of the flange is attributed to galvanic corrosion of the carbon steel material. Carbon steel is anodic to the adjacent alloy surfaces.

The pipe spool flange was replaced.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(A) as completion of a nuclear plant shutdown required by the plant's Technical Specifications.

05000336/LER-2011-00218 August 2011Millstone

At 1152 on June 20, 2011, with Millstone Power Station Unit 2 operating at 59 percent power in Mode 1, the reactor tripped automatically on low steam generator water level. The decreasing water level condition was due to a low suction pressure trip on the "B" steam generator feed pump (SGFP) that occurred while placing the "A" SGFP in service. The event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual or automatic actuation of any of the systems listed in 50.73(a)(2)(iv)(B). The actuation of the Auxiliary Feedwater Actuation System also is a reportable condition under the same paragraph.

The cause of the event was gaps in the application of operator fundamentals and some procedure quality issues associated with operations procedure OP 2204, Load Changes.

Procedure OP 2204 has been revised to ensure the second steam generator feedwater pump is placed in service at a lower power level. Additional corrective actions to address the underlying causes of the gaps in the application of operator fundamentals are being addressed in accordance with the station's corrective action program.,

05000336/LER-2011-00124 May 2011Millstone

On April 3, 2011, with Millstone Power Station Unit 2 (MPS2) in a refueling outage at 0% power in Mode 5, data taken during plant shutdown indicated that the Enclosure Building Filtration System had not met acceptance criteria rendering the Enclosure Building inoperable while MPS2 was in Mode 4. Plant Technical Specification (TS) 3.6.5.2 requires that the Enclosure Building shall be operable in Modes 1, 2, 3, and 4. The TS 3.6.5.2 Action requirements were met. Since the Enclosure Building did not meet the acceptance criteria, the safety function of the Enclosure Building to limit radiological releases in the event of a design basis accident could not be assured.

The direct cause for not meeting the Enclosure Building acceptance criteria was that sliding bushings on the main steam safety valves (MSSVs) exhaust piping had dislodged and not reseated. The apparent cause of this event was determined to be a design/application deficiency in the use of MSSV exhaust piping sliding bushings as an Enclosure Building boundary. A design change was implemented that no longer relies on the MSSV sliding bushings as Enclosure Building boundaries. Instead, improved boot seals located on the MSSV exhaust piping form the boundary for the MSSVs.

This condition is being reported pursuant to 10CFR50.73(a)(2)(v)(C) as an event or -condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.