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 Entered dateSiteRegionReactor typeEvent description
ENS 5382816 January 2019 08:12:00FitzPatrickNRC Region 1On January 16, 2019, with James A. Fitzpatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated that Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge while isolating Reactor Building Ventilation. The Secondary Containment differential pressure was less than 0.25 inches of vacuum water gauge for approximately ten (10) seconds, and then immediately returned to greater than or equal to 0.25 inches of vacuum water gauge. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector."
ENS 537785 December 2018 17:06:00FitzPatrickNRC Region 1At 1010 (EST) on December 5, 2018, Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge. This condition existed for approximately 3 minutes before the differential pressure was restored to normal when the Standby Gas Treatment system was manually initiated. This event was caused by a trip of the service air compressor 39AC-2A. The loss of instrument air pressure caused Reactor Building ventilation to isolate and raise Secondary Containment differential pressure. The instrument air pressure was restored when 39AC-2A was isolated and the two backup air compressors started. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee notified the NRC Resident Inspector.
ENS 526644 April 2017 11:32:00FitzPatrickNRC Region 1GE-4On April 4, 2017, at 0735 (EDT), the HPCI System was inadvertently isolated during the performance of l&C (Instrument and Control) testing. Technicians were in the process of performing instrument surveillance tests for the HPCI (high pressure coolant injection) System (using Allowed Out of Service Times) when a trip signal was applied to the incorrect instrument. This caused a HPCI System isolation signal on High Area Temperature, resulting in the closure of the HPCI steam isolation valves and rendering the system inoperable and unavailable. RCIC was immediately verified to be operable. The surveillance testing was aborted and system restoration is in progress. This condition is being reported as a condition that could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident per 10CFR50.72(b)(3)(v)(D). This placed the plant in a 14-day LCO action statement under Technical Specification 3.5.1. The licensee has notified the NRC Resident Inspector.
ENS 5264528 March 2017 14:47:00FitzPatrickNRC Region 1GE-4

The following report is made pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to an unintended initiation signal that occurred on January 31, 2017 with James A. FitzPatrick Nuclear Power Plant (JAF) in Mode 5 at zero (0) percent power. On January 31, 2017 at 1425 (EST) the control room received multiple annunciations associated with the following Systems / Trains: Primary Containment Isolation System (PCIS) / Trains A and B Residual Heat Removal System (RHR) / Trains A and B Core Spray (CS) / Trains A and B Reactor Core Isolation Cooling (RCIC) All four (4) Emergency Diesel Generators (EDG) auto-started with their associated Emergency Service Water pumps operating. RHR and CS both received initiation signals but were defeated per procedure. The HPCI (High Pressure Coolant Injection) auxiliary oil pump was taken to Pull-to-Lock per procedure, and the RCIC steam isolation valve cycled until the breaker was opened to close the valve. An evaluation concluded that the (Emergency Core Cooling System - ECCS) initiation signals were caused by the opening of a portable job box that was stored near sensitive equipment. Upon opening the job box, the lid bumped a reference leg resulting in the initiation signals. All initiation signals were reset and systems restored to normal shutdown lineups. The licensee notified the NRC Resident Inspector.

  • * * UPDATE ON 3/30/17 AT 0840 EDT FROM DUSTIN SCURLOCK TO DONG PARK * * *

To the original report, the licensee added, "This condition recurred at 1624 (EDT on 1/31/17). The licensee notified the NRC Resident Inspector. Notified R1DO (Cook).

ENS 5250322 January 2017 16:19:00FitzPatrickNRC Region 1GE-4Information from a Manual Phased Array UT (Ultrasonic Testing) examination of the 'A' RHR LPCI (Residual Heat Removal Low Pressure Coolant Injection) Injection Loop indicates an axially oriented indication 0.95 inch in length and 0.81 inch through wall. This is on weld number 24-10-130 (T to Valve dissimilar metal weld). This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A) based on the fact that the indications result in a defect in the primary coolant system which cannot be found acceptable under ASME Section XI. The licensee informed the NRC Resident Inspector. The weld is located where the RHR piping taps into the reactor vessel. The wall thickness at this location is 1.15 inches.
ENS 5249014 January 2017 13:09:00FitzPatrickNRC Region 1GE-4At 0613 EST on 1/14/2017, with the unit in Mode 2 at 0 percent power at the start of Refueling Outage 22, Drywell inspection identified a through-wall leak on the 3/4-inch vent line off the bonnet of valve 02MOV-43A, Reactor Water Recirc Pump A Suction Isolation Valve, in the Reactor Coolant System (RCS) loop inside the Primary Containment. This condition constitutes a defect in the primary coolant system. This event notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) as a condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The licensee has notified the NRC Resident Inspector.
ENS 5248010 January 2017 22:45:00FitzPatrickNRC Region 1GE-4This notification is a 10 CFR 21.21(a)(2) interim report for power supply model N-2ARPS-A6. Two instrument power supplies for the 'B' Residual Heat Removal (RHR) system were being bench tested prior to installation when it was discovered that they failed to meet Vendor Technical Manual specifications for voltage stability for varying loads. The deviation was a voltage drop of approximately 300mV. This did not meet the specification of less than 150mV when varying current from 5 amps (full load) to 2.5 amps. A second replacement power supply exhibited a similar 300 mV drop. James A. FitzPatrick (JAF) reviewed the work order instructions to determine if there was a deviation from the recommendations in the Foxboro technical manual F180-0309 Spec 200 Multinest Power Supply 2ARPS Series calibration. Since as-found voltage readings were within the required tolerance of the RHR instrument loops, the power supplies appear to have been capable to perform their intended function. However, this evaluation did not troubleshoot why the power supplies failed to meet the calibration requirements. The power supplies were sent to a repair vendor. The input from this vendor is expected to allow JAF to complete the evaluation per 10 CFR 21.21(a)(1) by March 21, 2017, and a notification for failure to comply or defect per 10 CFR 21.21(d)(3)(i) is expected by March 24, 2017, if necessary. This notification is being submitted as an interim report per 10 CFR 21.21(a)(2). The licensee notified the NRC Resident Inspector.
ENS 523432 November 2016 15:35:00FitzPatrickNRC Region 1GE-4

During panel walkdown, it was discovered that a tag out for the 'C' Residual Heat Removal pump suction valve was active and the valve was open with its breaker open. This rendered the valve inoperable and Technical Specification 3.6.1.3 Action C for penetration with one inoperable PCIVs was entered. The action was to isolate the penetration by closing the valve within (4) hours or restore power. The event was discovered at 0845 (EDT) and the breaker was closed at 0925 (EDT). Technical Specification 3.6.1.3 Action C (Isolate penetration within 4 hrs.) was entered at 0130 (EDT) (time breaker was opened per tagout) and exited at 0925 (EDT). This condition of non-compliance existed from 0530 (EDT) on 11/02/16 until 0925 (EDT) on 11/02/16. This event is being reported under 10CFR50.72 (b)(3)(v)(C). NRC Resident has been notified.

  • * * RETRACTION AT 1653 EST ON 1/3/2017 FROM MARK HAWES TO MARK ABRAMOVITZ * * *

In accordance with Technical Specification (TS) 3.6.1.3, Primary Containment Isolation Valves, the TS Basis states that one or more barriers are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. When two or more barriers are provided, one of these barriers may be a closed system. During this event, one of the barriers in the penetration became inoperable: 'C' Residual Heat Removal (RHR) pump suction valve 10MOV-13C. After the initial NRC notification, it was confirmed that the RHR system piping is classified as a closed system outside containment. The integrity of the closed-loop RHR system is verified by monitoring the keep-full system. Since the piping is maintained full of water during normal and post-accident modes of operation, a barrier against post-accident, gaseous, containment leakage is provided. Therefore, the affected penetration could have performed its intended safety function since there was redundant equipment in the same system which was operable. This event is not reportable under 10 CFR 50.72(b)(3)(v)(C) and the original notification may be retracted. Finally, the primary containment penetration with 10MOV-13C is with a closed system and the completion time per TS 3.6.1.3 Required Action C is 72 hours. The valve was restored to operable prior to exceeding this time. The licensee notified the NRC Resident Inspector. Notified the R1DO (Dentel).

ENS 5232928 October 2016 17:36:00FitzPatrickNRC Region 1GE-4

The licensee reported an unanalyzed condition under 10 CFR 50.72(b)(3)(ii)(B) due to a fire barrier/HELB (high energy line break) door being inoperable during maintenance. This resulted in two of five safe shutdown panels to be declared inoperable. The door, located between the Turbine and Administrative Buildings, was opened for approximately two minutes for 'tool pouch work'. When Operations discovered the door was opened for maintenance, they declared the door inoperable until Operations performed the surveillance required to declare the door operable. The total time the door was inoperable was approximately 1 hour and 11 minutes. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION AT 1642 EST ON 12/27/2016 FROM DUSTIN SCURLOCK TO MARK ABRAMOVITZ * * *

The condition reported in ENS 52329 pursuant to 10 CFR 50.72(b)(3)(ii)(B) has been evaluated, and determined not to be an unanalyzed condition that significantly degraded plant safety. NRC Regulatory Issue Summary (RIS) 2001-09, 'Control of Hazard Barriers,' allows breaching of HELB barriers, provided the risk associated with the applicable maintenance activity is assessed and managed in accordance with 10 CFR 50.65(a)(4) of the Maintenance Rule. The hazard barrier controls procedure at JAF (James A. Fitzpatrick) is consistent with this guidance, and includes compensatory measures for opening of the subject HELB door (76FDR-A-272-26). Per the JAF hazard barrier controls procedure the secondary HELB doors are to be verified operable, and the Alternate Shutdown Panels 25ASP-4 and 25ASP-5 declared inoperable. Based on a review of previous performances of ST-76Y, Fire Door Inspection and Operability Test, and the JAF Paperless Condition Reporting System, all applicable secondary HELB doors were operable prior to and during the 'tool pouch work' on 76FDR-A-272-26. JAF TS LCO 3.3.3.2, Remote Shutdown System (RSS), stipulates a completion time of thirty days to restore one or more required remote shutdown functions to operable. The duration of the 'tool pouch work' and inoperability of 76FDR-A-272-26 is well within this thirty day allowed outage time. In addition, the Alternate Shutdown Panels that were rendered inoperable by this condition are not required for mitigation of a HELB, and steam line break accidents are not discussed in the Technical Specification (TS) Bases for the Remote Shutdown System. The licensee notified the NRC Resident Inspector. Notified the R1DO (Lilliendahl).

ENS 5212225 July 2016 13:43:00FitzPatrickNRC Region 1GE-4A non-licensed supervisory employee had a confirmed positive test for alcohol during a for-cause fitness-for-duty test. The employee's access to the plant has been suspended. The licensee notified the NRC Resident Inspector.
ENS 520738 July 2016 08:48:00FitzPatrickNRC Region 1GE-4Oil reported in the vicinity of the station's circulating water system effluent after the start of 3rd circulating water pump. The source of the oil is believed to be from oil entrained in the discharge canal from oil leak previously reported in EN#52045. One circulating water pump was removed from service to mitigate the source. The United States Coast Guard Response Center, and the New York State Department of Environmental Conservation have been notified. James A. Fitzpatrick Control Room was notified of the issue at 0645, off site agencies were first notified at 0743. The licensee notified the NRC Resident Inspector. Notified DOE, EPA, USDA, HHS, and FEMA.
ENS 5204526 June 2016 23:08:00FitzPatrickNRC Region 1GE-4

The United States Coast Guard reported an oil sheen in the vicinity of the station's circulating water system effluent. Investigation by station personnel has not determined the source. The circulating water pumps were secured to mitigate the potential source. The United States Coast Guard response Center, and New York State Department of Environmental Conservation have been notified. The licensee notified the NRC Resident Inspector. Notified DOE, EPA, USDA, HHS, FEMA.

  • * * UPDATE ON 06/27/2016 AT 02:52 FROM DUSTIN SCURLOCK TO DAN LIVERMORE * * *

The source of the oil sheen has been identified. The source, main turbine lubricating oil, has been stopped and cleanup efforts are underway. Notified R1DO (Gray), DOE, EPA, USDA, HHS, and FEMA.

ENS 5204224 June 2016 16:06:00FitzPatrickNRC Region 1GE-4At 1215 (EDT) on 6/24/2016, James A. FitzPatrick (JAF) was at 100% power when Breaker 710340 tripped and power was lost to L-gears L13, L23, L33, and L43. These provide non-vital power to Reactor Building Ventilation (RBV), portions of Reactor Building Closed Loop Cooling (RBCLC), and 'A' Recirculation pump lube oil systems. Off-site AC power remains available to vital systems and Emergency Diesel Generators (EDG) are available. Due to the loss of RBV, Secondary Containment differential pressure increased. At 1215 (EDT), Secondary Containment differential pressure exceeded the Technical Specifications (TS) Surveillance Requirement SR-3.6.4.1.1 of greater than or equal to 0.25 inches of vacuum water gauge. The Standby Gas Treatment (SBGT) system was manually initiated and Secondary Containment differential pressure was restored by 1219 (EDT). The 'A' Recirculation pump tripped at 1215 (EDT) and reactor power decreased to approximately 50%. 'B' Recirculation pump temperature began to rise due to the degraded RBCLC system. At 1236 (EDT), a manual scram was initiated. Reactor Pressure Vessel (RPV) water level shrink during the scram resulted in a successful Group 2 isolation. All control rods have been inserted. The RPV water level is being maintained with the Feedwater System and pressure is being maintained by main steam line bypass valves. A cooldown is in progress and JAF will proceed to cold shutdown (Mode 4). Due to complete loss of RBCLC system, the Spent Fuel Pool (SFP) cooling capability is degraded but the Decay Heat Removal system remains available. SFP temperature is slowly rising and it is being monitored. The time (duration) to 200 degrees is approximately 117 hours. The initiation of reactor protection systems (RPS) due to the manual scram at critical power is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The general containment Group 2 isolations are reportable per 10 CFR 50.72(b)(3)(iv)(A). In addition, the temporary differential pressure change in Secondary Containment is reportable per 10 CFR 50.72(b)(3)(v)(C), as an event that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector and the State of New York.
ENS 519857 June 2016 13:07:00FitzPatrickNRC Region 1GE-4At 1030 EDT on 6/7/2016, both doors of a secondary containment airlock were reported to be simultaneously open for approximately 2 seconds during the normal passage of personnel. The brief time that the doors were simultaneously open constitutes an inoperable condition of secondary containment. Secondary containment differential pressure was maintained throughout the time period that the doors were open. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and per the guidance of NUREG-1022, Rev. 3, 'any event or condition that at the time of 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 licensee notified the NRC Resident Inspector.
ENS 5172210 February 2016 05:42:00FitzPatrickNRC Region 1GE-4The purpose of this report is to provide a telephone notification under 10 CFR 50.72(b)(2)(xi) to notify the NRC of the inadvertent actuation of one Oswego County notification siren at approximately 0247 (EST) on 02/10/2016. The James A. FitzPatrick Control Room was notified by Oswego County 911 at 0359 on 02/10/2016 of the inadvertent actuation. It is unknown at this time as to why the inadvertent alarm actuated. Siren repair personnel (ANS Services) have been dispatched to isolate the siren and begin repair work. The siren has since been silenced. Alternate notification of the public in the area is through Hyper Reach. The Oswego County Emergency Management Office issued a News Release identifying the inadvertent actuation of the emergency siren. The licensee will be notifying the NRC Resident Inspector.
ENS 5169429 January 2016 11:02:00FitzPatrickNRC Region 1GE-4

On January 29, 2016 James A. FitzPatrick Nuclear Power Plant (JAF) received notification from the site Sewage Treatment Plant (STP) operators that the January 6, 2016 monthly settleable solids result for the STP was 0.2 ml/L/hr (milliliter/liter/hour). This value exceeds the State Pollutant Discharge Elimination System (SPDES) permit limit of 0.1 ml/L/hr (daily maximum). The STP operators conduct daily process control tests at the STP and did not identify any system upset issues around the January 6, 2016 sample date, or any time since, that would be symptomatic of the slightly elevated settleable solids result. The JAF environmental engineer concluded that a notification to the New York State Department of Environmental Conservation (NYSDEC) was not required for this event; however, a courtesy notification for permit noncompliance was made. The NYSDEC has been notified. Pursuant to 10 CFR 50.72(b)(2)(xi), this condition is being reported as an event or situation for which notification to a government agency has been made. The NRC resident has been notified. JAF is currently at 0 percent power in Mode 2 following a forced outage resultant of events on January 23, 2016.

  • * * RETRACTION ON 1/29/16 AT 1630 EST FROM DUSTIN SCURLOCK TO DONG PARK * * *

Based on further review of the NRC reporting guidance relative to this criteria, JAF has concluded that this condition is below the reporting threshold outlined in NUREG-1022 Revision 3. NUREG-1022 states the following (page 54), 'Licensees generally do not have to report media and government interactions unless they are related to the radiological health and safety of the public or onsite personnel, or protection of the environment.' The condition originally reported in ENS 51694 is considered a minor deviation in sewage process limits, and has no impact on the radiological health and safety of the public or onsite personnel, or protection of the environment. Therefore, JAF is retracting ENS 51694. The NRC Resident Inspector has been notified. Notified R1DO (Bickett).

ENS 5168024 January 2016 01:10:00FitzPatrickNRC Region 1GE-4At 2241 (EST) on 1/23/2016, James A FitzPatrick inserted a manual scram from 89 percent power due to lowering intake level. Following the successful scram, a residual transfer occurred, resulting in a loss of the non-vital busses, loss of all Circulating Water Pumps, and a manual closure of the Main Steam Isolation Valves (MSIVs). The cause of the residual transfer is unknown. RPV (Reactor Pressure Vessel) level shrink during the scram resulted in a successful Group 2 isolation. Reactor Vessel (level) and pressure are being maintained with the High Pressure Coolant Injection System which was manually started. A cooldown is in progress. FitzPatrick will proceed to Mode 5 until the cause is identified and corrected. The Emergency Diesel generators auto started as a result of the loss of power to the non-vital busses. Offsite power remained available throughout the event. Operators are controlling pressure manually via the relief valves. FitzPatrick will notify the Public Service Commission of the event. The NRC Resident Inspector was notified.
ENS 5161319 December 2015 00:44:00FitzPatrickNRC Region 1GE-4On December 18, 2015 at 1722 EST, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, JAF received a notification pursuit to 10 CFR 21.21(d)(3)(ii) related to Moore Industries RTD temperature transmitters. Specifically, wire insulation in T2 transformer was damaged during assembly which reduced the insulation resistance and dielectric breakdown between the windings of the transformer. This equipment is in both redundant trains (A and B) of the Containment Atmosphere Dilution (CAD) System. Preliminary review by Operations and Engineering, which was completed on 12/18/15 at 2100 EST, determined the Part 21 results in both trains of CAD being inoperable and the applicable Technical Specification (TS) for both redundant trains of CAD being inoperable was entered. Per TS 3.6.3.2 Condition B, this places the unit in a 7-day shutdown LCO, provided the hydrogen control function is maintained. Per the TS Bases, the alternate hydrogen control capabilities are provided by the Primary Containment lnerting System, which is unaffected. The event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D), as an event or condition that could prevent fulfillment of a safety function. The licensee notified the NRC Resident Inspector.
ENS 515959 December 2015 22:28:00FitzPatrickNRC Region 1GE-4

During surveillance testing, 2 reactor recirculation loop flow transmitters, which input to the Average Power Range Neutron Monitors (APRM's) associated with the 'A' Reactor Protection Trip System (RPS), were found out of tolerance in the non-conservative direction. Non-conservative reactor recirculation flow setpoints for all 'A' side APRM's results in a loss of safety function for the APRM Neutron Flux High (Flow Biased) trip function of RPS. All instruments were adjusted back to within tolerance as allowed by the procedure, restoring the RPS safety function. Extent of condition and instrument drift issues are under evaluation via the corrective action process. This is an 8 hour reportable event under 10CFR50.72(b)(3)(v)(A) - Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shutdown the reactor and maintain it in a safe shutdown condition. The NRC Senior Resident Inspector has been notified.

  • * * RETRACTION ON 1/29/16 AT 1414 EST FROM DUSTIN SCURLOCK TO DONG PARK * * *

On the basis of a subsequent engineering evaluation, which reviewed the uncertainties considered within the setpoint calculations and the sequencing of the transmitter calibrations, it was determined that the APRM channels and Control Rod Block Monitor instrumentation associated with the Neutron Flux High (Flow Biased) function were unaffected by the out of calibration condition. Technical Specification (TS) Table 3.3.1.1 FUNCTION 2.b requires two (2) APRM Neutron Flux High (Flow Biased) channels per RPS trip system. TS Table 3.3.2.1 FUNCTION 1.a requires two (2) Rod Block Monitor (RBM) Upscale channels. These TS requirements were met upon discovery of this condition. The past-operability of the RPS and RBM instrumentation was unaffected by this condition. In addition, the engineering analysis confirmed that the Neutron Flux High (Flow Biased) allowable values in the Core Operating Limits Report (COLR) and the TRM were not exceeded. Therefore, there was no loss of safety function, and this condition was not reportable pursuant to 10 CFR 50.72(b)(3)(v)(A), as an event or condition that could have prevented fulfillment of a safety function. FitzPatrick is retracting ENS 51595. The degraded flow transmitters have been replaced, and the operability determination for the condition has been revised. All RPS and RBM instrumentation remain operable. The NRC Resident Inspector has been notified. Notified R1DO (Bickett).

ENS 515792 December 2015 14:14:00FitzPatrickNRC Region 1GE-4On December 1, 2015 at 2036 EST, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, Secondary Containment differential pressure exceeded the Technical Specification (TS) Surveillance Requirement (SR) of greater than or equal to 0.25 inches of vacuum water gauge for approximately one (1) minute and twenty (20) seconds. Secondary Containment (SC) had been declared inoperable prior to this event, to facilitate a planned evolution related to a previous failure that occurred on September 18, 2015 (reference EN #51409). Operators attempted to restore the Reactor Building Ventilation System (RBVS) to the normal system lineup upon completion of the planned evolution. The Secondary Containment differential pressure trended positive, and exceeded the TS SR differential pressure requirement during this transition. Preliminary investigations indicate that the cause of this event is associated with the Above Refuel Floor Exhaust Fan (66FN-13B). The design of the Above Refuel Floor Exhaust portion of the RBVS includes an interlock between the exhaust fan and a downstream damper position switch, which starts the fan when the damper is in the full open position. During the approximate one (1) minute and twenty (20) second duration that the TS SR was not met, 66FN-13B was not running with the associated discharge damper in the open position. Secondary Containment was operable after the SC differential pressure was restored upon start of 66FN-13B, and remains operable. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(c), as an event or condition that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector.
ENS 515123 November 2015 16:19:00FitzPatrickNRC Region 1GE-4On September 22, 2015, with James A. Fitzpatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated a spike in Secondary Containment differential pressure during performance of a surveillance test associated with automatic initiation of the Standby Gas Treatment System. Plant data systems recorded Secondary Containment differential pressure exceeding the Technical Specification allowed value. The Secondary Containment differential pressure was at or above zero inches of water for approximately ten (10) seconds, and then immediately trended negative following auto-start of one of the trains of Standby Gas Treatment. An operator was subsequently dispatched to the ventilation control panel, and verified that Secondary Containment differential pressure was more negative than the Technical Specification allowed value. This condition was entered into the Corrective Action Program, and subsequently, it was determined that the approximate ten second duration that Secondary Containment differential pressure was greater than the Technical Specification allowed value was reportable pursuant to 10 CFR 50.72(b)(3)(v)(C), as an event or condition that could have prevented fulfillment of a safety function. Secondary Containment was Operable following reestablishment of greater than or equal to 0.25 inches of water vacuum, and remains Operable. The licensee has notified the NRC Resident Inspector.
ENS 5148019 October 2015 08:19:00FitzPatrickNRC Region 1GE-4On 8/23/2015 at 1242 (EDT), with the reactor at 100% power, an invalid RPS MG (Reactor Protection System Motor-Generator) set 'A' trip resulting in a loss of RPS bus 'A'; this occurred during testing of the RPS instrument channels. All equipment operated as designed as a result of the loss of power to the 'A' RPS bus. The invalid trip was determined to be a result of the overvoltage relay being set too low. The above event meets the reporting criteria of 10CFR50.73(a)(2)(iv)(A) since the loss of RPS bus resulted in primary containment isolation signals affecting containment valves in more than one system. The following systems isolated as a result of the loss of 'A' RPS bus: Reactor Water Cleanup, Reactor Building ventilation, 'A' Containment Atmosphere Dilution, Torus Vent and Purge, Drywell Equipment and Floor Drain Sumps, 'A' Drywell Containment Atmospheric Monitors, Recirculation System Sample Line, Main Steam Line Drains and Residual Heat Removal drain valve to radwaste. 'A' Standby Gas Treatment System started as designed. This notification is being made in accordance with 10CFR50.73(a)(2)(iv)(A) to provide information pertaining to an invalid 'A' Reactor Protection System actuation. Completed actions were the replacement of overvoltage relay and voltage setpoint change, completed on 9/11/2015. In accordance with 10CFR50.73(a)(i) a telephone notification is being made instead of submitting a written Licensee Event Report.
ENS 5140918 September 2015 18:01:00FitzPatrickNRC Region 1GE-4At 1408 EDT on 9/18/2015, Secondary Containment Refuel Floor exhaust flow degraded due to an equipment malfunction in the running Refuel Floor exhaust train. The degraded exhaust flow caused Secondary Containment differential pressure to go positive for approximately three minutes, resulting in Secondary Containment being declared inoperable. Corrective action to start the Stand-by Gas Treatment System and the alternate Refuel Floor exhaust train restored Secondary Containment operable by re-establishing its required negative differential pressure. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and per the guidance of NUREG-1022, Rev. 3, 'Any event or condition that at the time of 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 Duty Team has been activated to develop a repair plan. The NRC Resident Inspector has been notified.
ENS 5140517 September 2015 16:24:00FitzPatrickNRC Region 1GE-4At 1220 (EDT) on 9/17/2015, both doors of a Secondary Containment airlock were reported to be simultaneously open for approximately five seconds during the normal passage of personnel. The brief time that the doors were simultaneously open constitutes an inoperable condition of Secondary Containment. Secondary Containment differential pressure was maintained throughout the time period that the doors were open. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and per the guidance of NUREG-1022, Rev. 3, 'Any event or condition that at the time of 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 NRC Resident Inspector has been notified.
ENS 5124220 July 2015 14:27:00FitzPatrickNRC Region 1GE-4On the morning of July 20, 2015 at 0740 EDT, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, the Secondary Containment differential pressure decreased below the JAF Technical Specification (TS) Surveillance Requirement (SR-3.6.4.1.1) value of greater than or equal to 0.25 inch of vacuum water gauge. Both trains of the Standby Gas Treatment System were placed in service and the Reactor Building was isolated. The decrease in Secondary Containment differential pressure was caused by Reactor Building roof maintenance creating multiple openings. Maintenance workers were immediately ordered to stop work and address the condition. Secondary Containment differential pressure was restored to within the TS SR value at 0915 EDT, and remains greater than 0.25 inch of vacuum water gauge. The secondary containment is a structure that surrounds the primary containment and is designed to provide secondary containment for postulated loss-of-coolant accidents inside the primary containment. To prevent exfiltration the secondary containment requires the control volume pressure at less than the external pressure. The differential pressure requirement of TS SR-3.6.4.1.1 ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration. During this period there were no unmonitored radioactive releases; however, this event could have prevented the fulfillment of a safety function to control the release of radioactive material and it is reported pursuant to 10 CFR 50.72(b)(3)(v)(C). The NRC Resident Inspector has been informed.
ENS 5115112 June 2015 14:05:00FitzPatrickNRC Region 1GE-4

At 1027 (EDT) on June 12, 2015, with the James A. FitzPatrick (JAF) Nuclear Power Plant operating at 100% reactor power, Oswego County Emergency Management Center notified JAF that the Tone Alert Radio System had been out of service since 0924 (EDT). This impacts the ability to readily notify a portion of the Emergency Planning Zone (EPZ) Population for the JAF Nuclear Power Plant. This failure meets NRC 8 hour reporting criterion 10 CFR 50.72(b)(3)(xiii). The county alert sirens which also function as part of the Public Prompt Notification System remain operable. The loss of the Tone Alert Radio System constitutes a significant loss of emergency off-site communications ability. Compensatory measures have been verified to be available should the Prompt Notification System be needed. This consists of utilizing the hyper reach system which is a reverse 911 feature available from the county 911 center. Local law enforcement personnel are also available for 'route alerting' of the affected areas of the EPZ. The event has been entered into the corrective action program and the (NRC) Resident Inspector has been briefed. National Weather Service is investigating the failures.

  • * * UPDATE FROM BENJAMIN EGNEW TO DONALD NORWOOD AT 1436 EDT ON 6/16/2015 * * *

The Tone Alert System was restored to service on 6/16/15 at 1230 EDT. The licensee notified the NRC Resident Inspector. Notified the R1DO (Bickett).

ENS 5115011 June 2015 23:37:00FitzPatrickNRC Region 1GE-4At 2205 (EDT) on June 11, 2015, with the James A. FitzPatrick (JAF) Nuclear Power Plant operating at 100% reactor power, Oswego County 911 Center notified JAF that the tone alert weather radios had been out of service since 2056 (EDT). This impacts the ability to readily notify a portion of the Emergency Planning Zone (EPZ) population for the JAF Nuclear Power Plant. This failure meets NRC 8 hour reporting criterion 10 CFR 50.72(b)(3)(xiii). The (Oswego) County alert sirens which also function as part of the public prompt notification system remain operable. The loss of the tone alert radios constitutes a significant loss of emergency off-site communications ability. Compensatory measures have been verified to be available should the prompt notification system be needed. This consists of utilizing the hyper reach system which is a reverse 911 feature available from the county 911 center. Local Law Enforcement personnel are also available for 'Route Alerting' of the affected areas of the EPZ. JAF was notified by Oswego County 911 Center that the tone alert radio system was restored to service at 2257 (EDT). The event has been entered into the corrective action program and the (NRC) Resident Inspector has been briefed.
ENS 5097912 April 2015 20:51:00FitzPatrickNRC Region 1GE-4

The purpose of this report is to provide a telephone notification under 10 CFR 50.72(b)(3)(v)(D) to notify the NRC of a temporary loss of the Control Room Envelope (CRE) boundary. The safety function of the CRE boundary is to ensure the in-leakage of unfiltered air into the CRE will not exceed the in-leakage assumed in the licensing basis analysis of design basis accident (DBA) consequences to CRE occupants. In addition to an intact CRE boundary maintaining CRE occupant dose from a large radioactive release below the calculated dose in the licensing basis consequence analysis for DBAs, it also ensures the occupants are protected from hazardous chemicals and smoke. The loss of the CRE boundary was due to a failed latching mechanism for a CRE boundary door used for normal passage of personnel into and out of the CRE. The failure of the door to latch as designed is considered a condition that could have prevented the fulfillment of a safety function at the time of discovery, and is therefore reportable as required by paragraph 50.72(b)(3), 'Eight-hour reports.' Procedural controls have restored the safety function of the CRE boundary by mechanically locking the subject door in the closed position through the use of a specifically designed mechanical strong-back until a permanent repair is made. The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM MARK HAWES TO JOHN SHOEMAKER AT 1642 EDT ON 6/1/15 * * *

The main control room corridor fire door (76FDR-A-300-10) was found to not be able to latch. The latch was stuck in the latch mechanism because the latch bolt was bent. The latch was replaced on 4/15/2015. The Control Room Emergency Ventilation Air Supply System (CREVAS) provides a protected environment from which occupants can control the plant following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. The Control Room Envelope (CRE) is the physical boundary around the CREVAS environment. The Operability of the CRE boundary depends on its ability to minimize in-leakage of unfiltered air such that after a design bases accident a habitable environment can be maintained for 31 days without exceeding 5 rem whole body dose or its equivalent to any part of the body. The control room is normally pressurized greater than the 0.125 inches water gauge. This causes air to leak out rather than allowing infiltration of air from surrounding areas into the CRE boundary. The pressurized control room pushes this door (76FDR-A-300-10) outward, toward the open direction; however, even though the latch to the door did not work the door was still able to close. The closed door minimized in-leakage and a positive differential pressure was maintained in the control room during this event. These doors are kept closed against the door seals primarily by the closure mechanism. The latch is a secondary means of ensuring that the doors remain closed as well as a means to control personnel access to the control room. The Control Room Envelope (CRE) remained Operable with this deficiency and there was no loss of safety function per 10 CFR 50.72(b)(3)(v)(D). The original notification may be retracted. The licensee has notified the NRC Resident Inspector. Notified the R1DO (Powell).

ENS 5057930 October 2014 13:05:00FitzPatrickNRC Region 1GE-4On the evening of October 28, 2014 at 1708 EDT, with James A. FitzPatrick (JAF) Nuclear Power Plant operating at 100 percent power, the Reactor Building differential pressure decreased below the JAF Technical Specification (TS) Surveillance Requirement (SR) value of at least 0.25 inches water vacuum for a period of thirty-four (34) seconds. This occurred during restoration of the Reactor Building Ventilation System (RBVS) following planned maintenance. The Reactor Building differential pressure was 0.50 inches water vacuum with the 'A' RBVS fans in-service in conjunction with the Standby Gas Treatment System (SGTS). The Reactor Building differential pressure decreased to 0.19 inches water vacuum when the SGTS was secured. The Reactor Building Vent was subsequently isolated, and the alternate 'B' RBVS fans were placed in-service; the differential pressure increased to within the required 0.25 inches water vacuum value. The JAF TS bases associated with Secondary Containment state that, 'for Secondary Containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.' Troubleshooting activities indicated that the transient was due to a non-safety related, non-TS damper downstream of one of the 'A' RBVS fans that did not fully stroke open. The subject damper is not part of Secondary Containment, and has no safety related function. This condition did not impact the leak tightness of Secondary Containment or the ability of the associated equipment to establish and maintain the required differential pressure. Secondary Containment would have fulfilled its safety function. However, because the JAF TS SR value of 0.25 inches water vacuum was not met, Secondary Containment was considered Technical Specification INOPERABLE for a period of thirty-four (34) seconds. The Secondary Containment is considered a single-train system; therefore, this condition is reportable pursuant to 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector.
ENS 5055822 October 2014 14:05:00FitzPatrickNRC Region 1GE-4A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been revoked.
ENS 5053214 October 2014 01:30:00FitzPatrickNRC Region 1GE-4

During the plant response to the trip of the B Recirculating water pump, reactor water level rose to the HPCI (High Pressure Core Injection) high water level trip setpoint as indicated on the associated instrumentation. With this high water level trip actuated, the HPCI high drywell pressure initiation signal would not have allowed the HPCI system to perform its intended safety function if required. If the HPCI system received the low water level initiation signal, the system would have been able to perform Its intended safety function. This high water level signal was actuated from 1935 (EDT) until reset at 1940 (EDT). This is reportable under 50.72(b)(3)(v). The licensee notified NRC Resident Inspector.

  • * * RETRACTION PROVIDED BY DAVID CALLAN TO JEFFREY HERRERA AT 1404 EDT ON 12/08/14 * * *

Further review has determined that the condition was not a result of procedural errors/inadequacies, equipment failures, or design / analysis inadequacies. Plant systems responded as per design when the HPCI system high water level trip actuated when reactor vessel water level rose to the HPCI high water level trip setpoint. HPCI initiation has two logics: one for low-low vessel water level and the other for a high drywell pressure. A vessel low-low water level is an indication that reactor coolant is being lost with a need for HPCI injection for core cooling. High drywell pressure could indicate a line break in the Reactor Coolant Pressure Boundary inside the drywell. The HPCI level instrumentation is designed to shut down the HPCI system upon high water level to prevent HPCI turbine damage due to gross moisture carryover and will re-initiate HPCI if vessel water level drops to the initiation water level setpoint. A HPCI high drywell pressure initiation signal, above setpoint, would have made up the logic for HPCI initiation and as per design, HPCI would have injected at the vessel low low level setpoint without operator action to reset the trip. In this instance, the trip was reset as prescribed by station procedures. HPCI was capable of performing its safety function after the high water level trip reset either by operator action or instrumentation (low low level initiation). The licensee will be notifying the NRC Resident Inspector. Notified R1DO (Rogge).

ENS 505094 October 2014 22:02:00FitzPatrickNRC Region 1GE-4

The total as-found Minimum Pathway Leakage Rate for the Primary Containment exceeded Level 1 acceptance criteria. Acceptance criteria of 321 (Standard Liters per Minute) SLM was not met. This criteria is equivalent to 1.0 La, the maximum allowable Primary Containment Leakage rate as prescribed by Technical Specification 5.5.6.c.1. This is reportable under 10CFR50.72(b)(3)(ii)(A) as 'The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded .. ' All other Level 1 acceptance criteria were met. All as-left containment leakage requirements for startup have been met. The licensee will notify the NRC Resident Inspector.

  • * * RETRACTION FROM DUSTIN SCURLOCK TO DANIEL MILLS AT 1646 EST ON 12/02/2014 * * *

On October 4, 2014, FitzPatrick reported that the total as-found containment minimum pathway leak rate exceeded the maximum allowable containment leak rate per the containment leakage rate testing program. This was primarily due to the drywell exhaust Penetration X26A/B. Penetration X26A/B Local Leak Rate Testing (LLRT) results were initially indeterminate, and therefore conservatively assumed to exceed the primary containment leakage acceptance criteria. The excessive leakage was assumed for Penetration X26A/B due to LLRT results for two (2) containment isolation valves (CIV). The subject CIVs are installed in series on Penetration X26A/B. The upstream valve is not isolable from primary containment, therefore, LLRT testing for these two CIVs is performed simultaneously via pressurization through a test connection between the two valves. During the LLRT, Penetration X26A/B was pressurized to 44.42 psig. The required test pressure for this penetration is 45.3 psig. As the required test pressure was not achieved, the LLRT results were initially indeterminate. Excessive leakage was conservatively assigned to the penetration resulting in the failure of the primary containment leakage acceptance criteria. This condition (failure of the primary containment leakage acceptance criteria) was determined to be reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) as a condition of the nuclear power plant, including its principle safety barriers, being seriously degraded. A subsequent engineering evaluation addressed the leakage for Penetration X26A/B, and concluded that the LLRT test results did not reflect failure of the primary containment leakage acceptance criteria. The installed configuration prevents testing these valves individually; however, troubleshooting activities indicated no detectable leakage through the downstream valve. The upstream valve was removed and inspected. The results of the inspection confirmed that all LLRT leakage was attributable to the upstream valve. Following maintenance activities, the valve was reinstalled and Penetration X26A/B was retested. The post-maintenance LLRT resulted in a total leakage of 0.078 SLM for Penetration X26A/B. The resultant total primary containment leakage rate determined on a minimum pathway basis was below the operability limits of 192 and 321 SLM (0.6 La and 1.0 La, respectively). Primary containment remained operable throughout Cycle 21; no degraded condition existed. Therefore, this (event notification) is being retracted. The licensee has notified the NRC Resident Inspector. Notified R1DO (Dentel)

ENS 5038921 August 2014 12:00:00FitzPatrickNRC Region 1GE-4

The purpose of this report is to provide a telephone notification under 10 CFR 50.72(b)(2)(xi) to notify the NRC of the inadvertent actuation of one Oswego County emergency notification siren at approximately 0850 (EDT) on 08/21/14. Thunderstorms in the area are believed to have caused a lightning strike and the spurious activation. Siren repair personnel have been dispatched to isolate the siren and begin repair work. The siren has since been silenced. Alternate notification of the public in the area is through Hyperreach. The Oswego County Emergency Management Office issued a news release identifying the inadvertent actuation of the emergency siren. The NRC Resident Inspector has been notified.

  • * * UPDATE AT 1145 EDT ON 8/23/2014 FROM HENK VERWAY TO MARK ABRAMOVITZ * * *

As of 0928 EDT on 08/23/2014, siren #13 has been repaired and returned to service. The licensee notified the NRC Resident Inspector. Notified the R1DO (McKinley).

ENS 5021018 June 2014 21:20:00FitzPatrickNRC Region 1GE-4

At 1545 (EDT), while testing of the Emergency Service Water system (ST-8Q) was being performed at the James A. FitzPatrick Nuclear Power Plant (JAF), two of five unit coolers (66UC-22H and 66UC-22K) in the East Crescent were found with indicated flow of 0 gpm. The other three unit coolers in the East Crescent Area were found with sufficient flow. At least four unit coolers are required to support the functionality of the East Crescent Area Ventilation Subsystem (TRO 3.7.C). The East and West Crescent Area Ventilation Subsystems support the Operability of the Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) system by removing heat from the areas, in the event that ECCS and RCIC were used to mitigate the consequences of an accident. The West Crescent Area Ventilation Subsystem remained functional. The accident mitigating function of the division of ECCS and RCIC located in the West Crescent Area were unaffected by this condition. However, this condition could have prevented the function of one division of the ECCS, including the single train of High Pressure Coolant Injection (HPCI), located in the East Crescent. Therefore, this condition could have prevented fulfillment of the safety function of HPCI and it is being reported under 10 CFR 50.72(b)(3)(v)(D). As part of the testing, the throttle valves to the unit coolers (66UC-22H and 66UC-22K) were cycled and normal flow was restored. This condition no longer exists. The licensee is investigating the loss of flow to the "H" and "K" unit coolers and the restoration of flow by cycling the unit cooler supply throttle valves. The licensee will be notifying the NRC Resident Inspector.

  • * * RETRACTION FROM DAVID CALLEN TO DANIEL MILLS AT 1506 EDT ON 8/13/2014 * * *

FitzPatrick is retracting EN # 50210 made on June 18, 2014 at 2120 EDT. The plant was at 86% power at the time. The ENS notification was an 8-Hr non-emergency notification to 10 CFR 50.72(b)(3)(v)(D) when it was discovered that two of five unit coolers in the East Crescent (66UC-22H and 66UC-22K) were found with indicated flow of 0 gpm while testing. The other three unit coolers in the East Crescent (66UC-22B, 66UC-22D, 66UC-22F) were found with sufficient flow. At least four unit coolers are required to support the functionality of the East Crescent Area Ventilation subsystem (TRO 3.7.C). The East and West Crescent Area Ventilation subsystems support the Operability of the Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) system by removing heat from these areas in the event that ECCS and RCIC are used to mitigate the consequences of an accident. As part of testing, throttle valves to unit coolers 66UC-22H and 66UC-22K were cycled and normal flow was restored. The West Crescent Area Ventilation subsystem remained functional. The accident mitigating function of the division of the ECCS and RCIC located in the West Crescent Area were unaffected by this condition. Initial review of this condition determined that it could have prevented the function of one division of the ECCS, including the single train of High Pressure Coolant Injection (HPCI), located in the East Crescent. Therefore, this condition was initially reported under 10 CFR 50.72 (b)(3)(v)(D) as a condition that could have prevented fulfillment of the Safety function of HPCI. This EN# 50210 is being retracted based upon a subsequent engineering analysis that determined that there is reasonable assurance that the three unit coolers with sufficient flow (66UC-22B, 66UC-22D, and 66UC-22F) would have been capable of removing accident heat loads as a function of time to maintain East Crescent area temperatures at a value which ensures operability of supported equipment. The analysis considered unit cooler heat transfer capability at the modified design condition flow of 22 gpm for historically observed lake temperatures and for flow at tested conditions. Additional margin in flow at the tested condition provided increased heat removal capability and provided added assurance that accident heat load would have been removed. The East Crescent Area Ventilation subsystem was, therefore, functional with three unit coolers (functionality never was lost) and the supported ECCS remained Operable. The Operability determination for the condition has subsequently been revised based upon the engineering analysis, to state the condition was not immediately reportable per 10 CFR 50.72. The licensee has notified the NRC Resident Inspector Notified R1DO (Kennedy)

ENS 501049 May 2014 16:46:00FitzPatrickNRC Region 1GE-4A non-licensed employee supervisor had a confirmed positive drug test during a random fitness-for-duty test. The employee's access to the plant has been revoked. The licensee will notify the NRC Resident Inspector.
ENS 499821 April 2014 14:02:00FitzPatrickNRC Region 1GE-4

At 0645 EDT on the morning of April 1, 2014, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, the Control Room received an alarm associated with the ventilation system for the 'B' division of the Residual Heat Removal Service Water (RHRSW) and Emergency Service Water (ESW) pump room. Investigation identified that the ventilation exhaust fan (73FN-3B) associated with this pump room had tripped due to thermal overload. The overload relay was reset at 0704 EDT and the fan automatically started; the fan is currently operable. During this period, the fan would not have automatically started. The ventilation systems for the RHRSW and ESW pump rooms are not included in the JAF Technical Specifications (TS), nor are they in the JAF Technical Requirements Manual (TRM). The ambient temperature limit in the RHRSW and ESW pump room was never challenged. However, with 73FN-3B non-functional, it is procedurally required to declare 10P-1B (RHRSW Pump B), 10P-1D (RHRSW Pump D) and 46P-2B (ESW Pump B) inoperable. The 'B' ESW pump cools the 'B' EDG subsystem, which would therefore also be inoperable. During this period, the 'A' EDG subsystem was inoperable for an emergent issue. Because the 'A' and 'B' EDG subsystems were concurrently inoperable for a period of approximately 45 minutes, this condition resulted in a loss of safety function for the Emergency Diesel Generators, which is reportable pursuant to 10 CFR 50.72(b)(3)(v)(A). Both affected emergency diesel generators were in the same division with another redundant division operable. The licensee notified the NRC Resident Inspector and will be notifying the State of New York.

  • * * UPDATE AT 1200 EDT ON 4/2/14 FROM CHRIS ADNER TO S. SANDIN * * *

The statement "Both affected emergency diesel generators were in the same division with another redundant division operable" is incorrect. Both divisions of emergency diesel generators were inoperable, since one of two available emergency diesel generators per division were inoperable at the time of this event. The Licensee notified the NRC Resident Inspector. Notified R1DO (Cahill).

ENS 4994121 March 2014 12:53:00FitzPatrickNRC Region 1GE-4A significant Fitness for Duty (FFD) programmatic vulnerability was discovered during an NRC inspection. A potential exists that some members of the FFD site's random drug test pool could control and predict the date and time that the random list is run; thus mitigating the effectiveness of the 'random aspect' for the staff. Per 10 CFR 26.31(d)(2)(i), 'Random testing. Random testing must - Be administered in a manner that provides reasonable assurance that individuals are unable to predict the time periods during which specimens will be collected.' Since, these individuals may be able to predict the timeliness of random drug test events this condition is reportable per 10 CFR 26.719(b)(4) as a discovered programmatic vulnerability of the FFD program that may permit undetected drug or alcohol use by individuals who are assigned to perform duties that require them to be subject to the FFD program. The licensee has notified the NRC Resident Inspector.
ENS 497945 February 2014 15:15:00FitzPatrickNRC Region 1GE-4At 1130 (EST) on February 5, 2014, with the James A. FitzPatrick Nuclear Power Plant (JAF) at 100% reactor power, Oswego County Emergency Management notified JAF that the tone alert radios had been out of service since 1000. This impacts the ability to readily notify a portion of the Emergency Planning Zone (EPZ) population for the Nine Mile Point and JAF nuclear power plants. This failure meets NRC 8-hour reporting criterion 10 CFR 50.72(b)(3)(xiii). The county alert sirens, which also function as part of the public prompt notification system, remain operable. The loss of the tone alert radios constitutes a significant loss of emergency off-site communications ability. Compensatory measures have been verified to be available should the prompt notification system be needed. This consists of utilizing the hyper reach system which is a reverse 911 feature available from the county 911 center. Local law enforcement personnel are also available for route alerting of the affected areas of the EPZ. At 1328 on February 5, 2014, JAF was notified by Oswego County Emergency Management that the tone alert radios had been returned to service at 1325. The licensee notified the NRC Resident Inspector, the State of New York and local authorities of the outage.
ENS 4966018 December 2013 18:56:00FitzPatrickNRC Region 1GE-4During the performance of surveillance testing on 12/17/2013 for the HPCI CST low water level switch instrument functional test, the safety function for the low CST level suction swap-over was lost. While performing testing, 23LS-75B setpoint was found to be below the TS required value with 23LS-74B already declared inoperable. Tech Spec 3.3.5.1 Condition D was entered and within 1 hour of finding 23LS-75B setpoint below the TS required value, this switch was adjusted to within tolerance as allowed by procedure and tested satisfactory. Both switches were inoperable concurrently for a period of less than 1 hour from time of discovery of 23LS-75B being out of tolerance. This issue has been discussed with the NRC Resident Inspector.
ENS 4955118 November 2013 12:08:00FitzPatrickNRC Region 1GE-4James A. FitzPatrick Nuclear Power Plant (JAF) was notified at 0512 EST by the New York State (NYS) Watch Center that the Radiological Emergency Communications System (RECS) and commercial telephones were not available. The unavailability of the communications systems was a result of an unplanned computer server outage affecting the New York State Watch Center. While the RECS line remained operational, it was not available due to the relocation of personnel from the NYS Watch Center to an alternate location. An alternate method of communication was established via cell phone at the time of notification. This condition is reportable as a major loss of emergency offsite communications capability under 10 CFR 50.72(b)(3)(xiii). The NYS Watch Center network and communications systems have been restored and the facility staffed as of 0757 EST. The condition has been entered into the station's corrective action program. The NRC Resident Inspector has been notified.
ENS 4949131 October 2013 20:41:00FitzPatrickNRC Region 1GE-4Per review of OE INPO ICES-305419, 'Unfused remote DC ammeter circuit could result in a secondary fire due to multiple fire induced faults' from Davis-Besse and Cooper Condition Report CR CNS-2013-07413; it has been determined that JAF (James A. Fitzpatrick) is potentially susceptible to the same condition. The condition in the Davis-Besse OE is described as follows: The wiring design for the ammeters contains a shunt in the current flow from each direct current (DC) battery or charger. Bolted on the shunt bar are two IEEE 383 qualified leads to a current meter in the main control room (MCR). The small difference in voltage between the two taps on the shunt is enough to deflect the current gauge in the MCR when current flows from the battery or charger through the shunt. The ammeter wiring attached to the shunt does not have fuses. It is postulated that a fire could cause one of these ammeter wires to short to ground at the same time the fire causes another DC wire from the opposite polarity on the same battery to also short to ground. This would cause a ground loop through the unfused ammeter cable. With enough current going through the cable, the potential exists that the cable could self-heat to the point of causing a secondary fire in the electrical tray at some point along the path of the cable (including the Control Room) or possibly heat up to the point of causing damage to adjacent cables that may be required for safe shutdown. TRM 3.7.M, Fire Barrier Penetrations, is applicable. The functional integrity of the fire barrier penetration seals ensures that fires will be confined or adequately retarded from spreading to adjacent portions of the facility. This design feature minimizes the possibility of a single fire rapidly involving several areas of the facility prior to detection and extinguishment. The fire barrier penetration seals are a passive element in the facility fire protection program and are subject to periodic inspections. The issue identified is with the potential of a fire starting in another location other than the original fire location bypassing fire barriers due to a fire induced electrical short. Per engineering, the areas with the deficient fire barriers are the DC Switchgear Rooms A and B, Cable Spreading Room, Relay Room and Control Room. An active LCO will track the TRM action for the non-functionality of those fire barriers. This condition is being reported under 10CFR50.72 (b)(3)(ii)(B) as a condition that results in the plant being in an unanalyzed condition which significantly degrades plant safety. The licensee has notified the NRC Resident Inspector.
ENS 491798 July 2013 14:11:00FitzPatrickNRC Region 1GE-4This 60-day telephone notification is provided in accordance with 10 CFR 50.73(a)(1) to report an invalid actuation of the Emergency Diesel Generators (EDGs) reportable under 10 CFR 50.73(a)(2)(iv)(A). On the morning of May 10th, 2013, at approximately 0141 EDT the B Emergency Diesel Generator (EDG) subsystem was inoperable for a planned maintenance work window on the B EDG only (there are two EDGs per subsystem). The work was completed to the point that protective tagging removal and testing of the B EDG were the only activities left. During the process of removing protective tagging, an operator incorrectly opened a switch-gear cubicle door that contained the potential transformer fuses for the B division emergency electrical bus (10600 bus); the system is designed such that opening the cubicle door disconnects the fuses from the circuit. The potential transformers reduce the associated bus voltage (in this case the 10600 bus), nominally 4160V, to 120V for the monitoring and control circuits. These circuits monitor the 10600 bus voltage and initiate a start signal for the EDGs should voltage be degraded or lost. After a time delay of less than or equal to 2.5 seconds, if voltage is not restored, the normal supply breaker for the 10600 bus will open and the EDG output breakers will close re-powering the bus from its emergency source. The response of the 10600 bus control and monitoring circuits functioned as designed. An initiation signal was sent to the D EDG control system when the monitoring circuit sensed no bus voltage (potential transformer fuses removed; door open). With the immediate closure of the cubicle door, the monitoring circuit was restored and 10600 bus voltage was sensed. This resulted in the under voltage logic returning to normal hence, the 10600 bus did not disconnect from its normal station power source and the D EDG output breaker did not close to power the 10600 bus. Subsequently the D EDG was secured (0146 EDT on May 10th) and protective tagging removal and testing of B EDG continued. The B EDG subsystem was declared operable on May 12, 2013 at 0315 EDT. This event did not result in any adverse impact to the health and safety of the public. The licensee has notified the NRC Resident Inspector.
ENS 4907026 May 2013 21:52:00FitzPatrickNRC Region 1GE-4At 1839 EDT on May 26, 2013, with the James A. FitzPatrick (JAF) Nuclear Power Plant performing power ascension from 98% reactor power, Oswego County Emergency Management notified JAF that the Tone Alert Radios had been out of service since 1745 EDT. This impacts the ability to readily notify a portion of the Emergency Planning Zone (EPZ) Population for the Nine Mile Point and JAF Nuclear Power Plants. This failure meets NRC 8-hour reporting criterion 10 CFR 50.72(b)(3)(xiii). The County Alert Sirens which also function as part of the Public Prompt Notification System remain operable. The loss of the Tone Alert Radios constitutes a significant loss of emergency off-site communications ability. Compensatory measures have been verified to be available should the Prompt Notification System be needed. This consists of utilizing the hyper reach system which is a reverse 911 feature available from the county 911 center. Local Law Enforcement Personnel are also available for 'Route Alerting' of the affected areas of the EPZ. At 2111 EDT May 26, 2013, JAF was notified by Oswego County Emergency Management that the Tone Alert Radios had been returned to service The event has been entered into the corrective action program and the NRC Resident Inspector has been briefed.
ENS 4898129 April 2013 17:50:00FitzPatrickNRC Region 1GE-4Offsite notification to the New York DEC (Department of Environmental Conservation) to report a Freon (R-22) release to the air of 8 lbs. 11 ozs. This release came from the cafeteria kitchen walk-in cooler. The licensee notified the NRC Resident Inspector.
ENS 4895823 April 2013 11:13:00FitzPatrickNRC Region 1GE-4James A. FitzPatrick Nuclear Power Plant (JAF NPP) was notified by NYS Warning Point that the RECS (Radiological Emergency Communication System) line and all land lines to NYS were non-functional beginning at approximately 0330 (EDT) on 4/23/13. Due to this condition, JAF NPP did not have any communications with the NYS Warning Point available via NORMAL or BACKUP methods per Emergency Plan Procedures. At 0449 (EDT) on 4/23/13, the JAF NPP control room was provided with an alternate means of contacting the NYS Warning Point via cell phone and hence a viable means of BACK-UP communications was established. Per 10 CFR 50.72 (b)(3)(xiii), 'any event that results in a major loss of off-site communications capability (off-site notification system between licensee and off-site officials - NYS) is reportable via 8-hour report.' Subsequent to this event at approximately 0930 (EDT) 4/23/13, functionality of the RECS was restored and tested satisfactorily. The licensee notified the NRC Resident Inspector as well as the State and local governments.
ENS 4892716 April 2013 10:17:00FitzPatrickNRC Region 1GE-4The purpose of this report is to provide a telephone notification under 10CFR50.72(b)(2)(xi) to notify the NRC of the inadvertent actuation of the Oswego County emergency notification sirens at approximately 0845 (EDT) on 4/16/13. Oswego County was performing routine weekly testing and siren #17 was inadvertently actuated for approximately 2 minutes. The Oswego County Emergency Management Office issued a News Release identifying the inadvertent actuation of the emergency siren. The NRC Resident Inspector has been notified.
ENS 4890710 April 2013 11:06:00FitzPatrickNRC Region 1GE-4

A planned work package (52343687-01) at the James A. FitzPatrick (JAF) Nuclear Power Plant will be performed for DOP/Freon Testing TSCVASS as required per TS 5.5.8 Ventilation Filter Testing Program. The testing requires breaking the boundary into the Technical Support Center (TSC) ventilation system to obtain a charcoal sample. Therefore, the TSC ventilation system will be rendered nonfunctional during the duration of this work activity. The TSC ventilation is expected to be out of service for approximately 6 hours. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the station Emergency Plant Manager will relocate the TSC staff to an alternate TSC location In accordance with applicable site procedures giving first consideration to the Control Room. TSC facility leads have been made aware of this contingency. This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the potential loss of an emergency response facility (ERF). An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 4/10/13 AT 1717 EDT FROM DAVE RICHARDSON TO DONG PARK * * *

This is an update from EN #48907. Planned maintenance has been completed on the Technical Support Center (TSC) ventilation system. The TSC filtered ventilation system has been restored to normal standby lineup. The NRC Resident Inspector has been informed.

ENS 4867616 January 2013 04:43:00FitzPatrickNRC Region 1GE-4On January 15, 2013 at 2112 EST while performing testing associated with the remote shutdown system at the James A. FitzPatrick Nuclear Power Plant, an unexpected loss of the 10600 bus 'B' division AC vital power system occurred. This loss of power to the 10600 bus resulted in an automatic actuation of the 'B' and 'D' Emergency Diesel Generators. The diesel generators started as expected, but did not close in to energize the 10600 Bus due to the configuration at the time of the event. As a result of the loss of the 10600 bus, the 'B' Reactor Protection System (RPS) lost power resulting in a half scram signal and a Group II Primary Containment Isolation System (PCIS) actuation. This actuation resulted in closing containment isolation valves in multiple systems and isolating Reactor Water Clean-Up (RWCU). Based on these system actuations, the event is reportable under criterion 10 CFR 50.72(b)(3)(iv). Power to the 10600 Bus was restored January 16, 2013 at 0400 EST, and the half scram and isolation signals have been reset. Additional actions to restore systems to a normal operating line-up are on-going. Investigation into the cause of the unexpected power loss is on-going and will be addressed through the corrective action program. The NRC Resident Inspector has been notified.
ENS 486547 January 2013 14:59:00FitzPatrickNRC Region 1GE-4

A planned work package (52232185-01) at the James A. FitzPatrick (JAF) Nuclear Power Plant will be performed for inspection and verification of 72AOD-171 (Admin Building Ventilation AHU-4 Fresh Air Supply Isolation Damper). The isolations necessary to perform this evolution will require tagging out 72AHU-4 (Admin Building Office Area Air Handling Unit) and breaking the boundary into the Technical Support Center (TSC) ventilation system. Therefore, the TSC ventilation system will be rendered non-functional during the duration of this work activity. The TSC ventilation is expected to be out of service for approximately 10 hours. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the station Emergency Plant Manager will relocate the TSC staff to an alternate TSC location in accordance with applicable site procedures giving first consideration to the Control Room. TSC facility leads have been made aware of this contingency. This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the potential loss of an emergency response facility (ERF). An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 01/08/13 AT 1654 EST FROM BOB WISE TO HUFFMAN * * *

The TSC ventilation system was returned to service at 1650 EST. The NRC Resident Inspector has been notified. R1DO (Newport) notified.

ENS 4858112 December 2012 12:30:00FitzPatrickNRC Region 1GE-4Event Summary: On October 13, 2012, at approximately 0218 (EST), a full reactor scram signal was received in the James A. FitzPatrick (JAF) control room. At the time of this event, the plant was in cold shutdown (Mode 4) and refueling outage 20 (R20) was in progress. The scram signal occurred because Reactor Vessel Scram & Primary Containment Isolation Level Transmitter (02-3LT-101C) and Reactor Vessel Scram & Primary Containment Isolation Level Transmitter EQ (02-3LT-101D) momentarily failed downscale, and then immediately recovered. 02-3LT-101C is an 'A' division component and 02-3LT-101D is a 'B' division component. Therefore, both the 'A' and 'B' divisions of reactor protection actuated providing a full reactor scram and outboard primary containment isolation signal. AOP-15, 'Isolation Verification and Recovery' verified that the proper containment isolation response was received. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) because it resulted in the invalid actuation of the reactor protection system while the reactor was already shut down. Apparent Cause: A Failure Mode Analysis and Apparent Cause Evaluation were performed to determine the most likely cause of both level transmitters to momentarily spike downscale. It was concluded that the most probable cause was due to a worker inadvertently coming into contact with the level transmitters' exposed sensing lines. Contributing to this event was the misjudgment by the Operations individual reviewing the work package, on the risk significance of the instrumentation in the vicinity of the work area. As a result, the work area was not constructed in a manner to preclude interference with the level transmitters or associated sensing lines. Corrective Actions: Immediate corrective actions were to walk down the affected instrument lines to ensure no damage had been caused. Additional corrective actions were to install signs near the level transmitters and exposed sensing lines. The signs denote that sensitive instrument lines are present. Future corrective actions include a walk down by engineering and operations to identify other areas in the plant where sensitive instrument lines are present and place additional signs or barriers as appropriate. The licensee notified the NRC Resident Inspector.