Semantic search
Entered date | Site | Region | Reactor type | Event description | |
---|---|---|---|---|---|
ENS 57437 | 22 November 2024 12:50:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: This notification is a 10 CFR 21.21(a)(2) interim report for General Electric thermal overload relay, model CF124G011, part number DD317A7861P003. A sample of overload relays were sent to PowerLabs for parts quality initiative testing. The results were reviewed by James A. FitzPatrick Nuclear Power Plant (JAF) and a deviation in one relay component was discovered. Testing identified a failure to latch on trip, which is a deviation from the performance characteristics of the relay. Under normal operation, the relay would latch in the tripped state requiring a manual reset of the relay. If the relay with the deviation were installed, the relay would trip when required; however, it would automatically reset. The unexpected reset could result in unintended cycling of associated equipment including repeated exposure to inrush current and potential damage. Bench testing would be expected to identify this condition prior to installation. Based on a review, this potential condition does not affect installed equipment. The affected relay was stored at JAF since July 1998. The cause of the deviation cannot be investigated because the part is not available; however, the evaluation of the potential effect of the condition on equipment where the relay could have been used at JAF is ongoing, and it is expected to be completed by February 28, 2025. This notification is being submitted as an interim report per 10CFR21.21(a)(2). The NRC resident inspector has been notified. |
ENS 57333 | 23 September 2024 11:02:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: At 0720 EDT on September 23, 2024, James A. FitzPatrick was at 100 percent power when an automatic scram occurred as a result of a main turbine trip due to an automatic trip of the generator output breakers; the cause is still under investigation. The scram was not complex. The automatic scram inserted all control rods. A subsequent reactor pressure vessel (RPV) low water level resulted in a group 2 isolation and initiation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems. RCIC did inject, but HPCI did not inject, as expected, based on RPV water level recovery with the feedwater system. Reactor pressure is being maintained by main steam line bypass valves. The plant is stable in Mode 3 with the 'A' reactor feed pump maintaining RPV water level. The initiation of the reactor protection system (RPS) due to the automatic scram signal while critical is reportable per 10 CFR 50.72(b)(2)(iv)(B). The general containment Group 2 isolations and HPCI and RCIC system actuations are reportable per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The group 2 containment isolation affects multiple systems.
On 9/23/2024 at 1156 EDT Constellation communications provided a media statement to Oswego area news media contacts summarizing the events that had occurred at Nine Mile Point Unit 2 and FitzPatrick Unit 1. This is a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified. Notified R1DO (Dimitriadis) |
ENS 57172 | 13 June 2024 17:12:00 | FitzPatrick | NRC Region 1 | GE-4 | At 1331 EDT on 6/13/2024, it was determined that a non-active licensed operator supervisor tested positive in accordance with the fitness for duty testing program. The individual's authorization for site access has been denied. The NRC Resident Inspector has been notified. |
ENS 57120 | 9 May 2024 21:10:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: At 1629 EDT on 05/09/2024, the high pressure coolant injection (HPCI) system was declared inoperable due to a pinhole through-wall leak identified on the seal drain line for 23HOV-1 (HPCI trip throttle valve) downstream of the restricting orifice 23RO-137A. The location of the defect is in the class 2 safety related piping. HPCI is a single train safety system and this notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: This pinhole leak was discovered during normal operator rounds. Although HPCI is declared inoperable and in a 14-day limited condition of operation, the system function remains available. In addition, all other ECCS systems are currently operable. Compensatory measures (walkdowns) have been implemented to ensure the leak rate does not significantly increase.
FitzPatrick performed an additional technical evaluation of the steam leak identified on May 9, 2024. The evaluation concluded that the HPCI system would have remained operable and performed its specified safety function with a postulated complete failure of this pipe, considering its size, location, and impact of the leak. Additionally, all components in the vicinity would have retained their required safety functions. Based on this conclusion, EN 57120 is being retracted. The NRC Senior Resident Inspector has been notified. Notified R1DO (Elkhiamy). |
ENS 57117 | 9 May 2024 14:42:00 | FitzPatrick | NRC Region 1 | The following is a synopsis of information provided by Engine Systems, Inc. (ESI) via fax and email: On February 22, 2024, an EMD brand cylinder liner developed a jacket water leak following installation on an emergency diesel generator set. The leak occurred at a brazed joint and was detected after post-installation engine testing. Had the leak gone undetected, jacket water may have accumulated in the combustion chamber or airbox and potentially contaminated the engines lubricating oil. Jacket water intrusion into any of these areas is undesirable and could lead to failure of the diesel engine and therefore failure of the emergency diesel generator set. The extent of condition is a single cylinder liner, P/N 9318833, S/N 20D6294 used in the power assembly shown below. Customer: Constellation - Fitzpatrick Customer PO: 703, release 13498 ESI Sales Order: 3021545 Part Number Ordered: 40124898 (Blade Power Pack) Serial Number: 20L0603 ESI C-of-C Date: April 1, 2021 The corrective action: For Fitzpatrick: No action required; the power assembly has been returned to ESI for replacement. For ESI: ESI will revise the dedication package to include additional verifications to prevent reoccurrence. The revision will be implemented within 30 days. Name and contact information: Dan Roberts, Quality Manager Engine Systems Inc. 175 Freight Rd. Rocky Mount, NC 27804 John Kriesel, Engineering Manager Engine Systems Inc. 175 Freight Rd. Rocky Mount, NC 27804 | |
ENS 57072 | 12 April 2024 11:49:00 | FitzPatrick | NRC Region 1 | The following is a summary of the information provided by Engine Systems Inc. (ESI) via facsimile: An EMD (Brand Name: Electro-Motive Diesel) cylinder liner developed a jacket water leak following installation on an emergency diesel generator set at the James A. Fitzpatrick Nuclear Power Plant. The leak occurred at a brazed joint and was detected after post-installation engine testing. Had the leak gone undetected, jacket water may have accumulated in the combustion chamber, airbox, and/or lubricating oil which could have eventually led to failure of the emergency diesel generator set. ESI was the supplier of the EMD cylinder liner (part number: 9318833, serial number: 20D6294). The EMD cylinder was a component of a Blade Power Pack Assembly, part number: 40124898, serial number: 20L0603 Corrective Actions: ESI will revise the dedication package to include additional verifications to prevent reoccurrence. The revision will be implemented within 30 days. Fitzpatrick returned the power assembly to ESI for replacement and no further action is required from Fitzpatrick. Affected Plants: Fitzpatrick. No other sites known to be affected. The name and address of the individuals reporting this information is: John Kriesel Engineering Manager Engine Systems, Inc.; 175 Freight Rd. Rocky Mount, NC 27804 Dan Roberts Quality Manager Engine Systems, Inc.; 175 Freight Rd. Rocky Mount, NC 27804 | |
ENS 56822 | 30 October 2023 17:06:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone call and email: A non-licensed supervisory employee had a confirmed positive test during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. |
ENS 56667 | 8 August 2023 17:03:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: A licensed (non-active) individual failed to comply with fitness for duty testing policies. The individual's unescorted access was terminated. |
ENS 56373 | 19 February 2023 08:56:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via fax or email: At 0105 EST on February 19, 2023, with the James A. FitzPatrick Nuclear Power Plant (JAF) at 100 percent power, a valid high main steam line radiation signal was received. An actuation of a fire protection foam system caused migration of high conductivity water into a low conductivity sump. Organic compounds were introduced into the primary coolant and resulted in a temporary increase in nitrogen-16 which was detected by main steam line radiation monitors and actuated primary containment isolation signals in more than one system. The reactor water recirculation sample system isolated. The signal also went to the normally isolated main steam line drain system and condenser air removal system. The event is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). The elevated radiation condition no longer exists. Health and safety of the public was not impacted by this event. The NRC Resident Inspector was notified. |
ENS 56362 | 6 February 2023 13:26:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by Constellation via email: On 02/06/2023 at 0416 EST, the Constellation Emergency Response Organization (ERO) Notification Database System uploaded data files into the Mass Notification System (Everbridge) which is used to notify ERO personnel when activated. At 0630, the individual reviewing the uploaded files discovered that the data files did not upload properly and that Everbridge may not notify all ERO individuals within the required 10 minutes of system initiation. Constellation resolved the issue by 0752. During the time period of 0416 to 0752, control room operators would have been unaware that the ERO notification was not successful. Therefore, this issue constitutes a loss of offsite communications capability and is reportable under 10 CFR 50.72(b)(3)(xiii), 'The licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' This loss of offsite communications capability affected all Constellation nuclear stations. There was no impact on the health and safety of the public or plant personnel. Each affected station NRC Resident Inspectors have been or will be notified. |
ENS 56158 | 12 October 2022 09:21:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: On 10/08/22, a non-supervisory employee violated the station's FFD policy. The individual's site access has been terminated. |
ENS 56119 | 26 September 2022 05:41:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0001 EDT on September 26, 2022, James A. FitzPatrick (JAF) removed the generator from service as part of a planned shutdown for refueling. At 0306 EDT, with the mode switch in Startup/Hot Standby and inserting rods, JAF experienced a spurious Scram and closure of seven out of eight main steam isolation valves (MSIV's). The reactor protection system (RPS) actuated during the event, resulting in all control rods being fully inserted. The cause of the closure of MSIV's and the Scram is being investigated. This condition is being reported as a four-hour NRC report per 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation, and as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the safety system actuation based on the multiple main steam isolation valves closing on an isolation signal. There was no impact to the health and safety of the public. The NRC Resident Inspector has been notified.
The following update was provided by the licensee via email: This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A) for Reactor Protection System (RPS) actuation along with Main Steam Isolation Valves (MSIV) system actuation. An analysis of reactor criticality was performed for the period of time prior to the RPS actuation event. Operators were inserting control rods per the shutdown Reactivity Management Plan. The Intermediate Radiation Monitoring (IRM) readings preceding the scram signal demonstrate a negative reactivity direction without control rod movement. The analysis concluded that the reactor was subcritical when RPS was actuated. The NRC Resident Inspector has been notified. Notified R1DO (Young). |
ENS 55871 | 29 April 2022 20:44:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 1251 EDT on April 29, 2022, while troubleshooting the failure of the High Pressure Coolant Injection (HPCI) Exhaust Drain Pot High Level Alarm to clear, it was discovered that the High Pressure Coolant Injection exhaust line condensate drain system was not functioning as designed to support removal of condensate from the turbine exhaust. This resulted in some water accumulation in the turbine casing. Subsequently, the High Pressure Coolant Injection System was declared inoperable. As a result, this condition is being reported under 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented fulfillment of the safety function at the time of discovery.
A technical evaluation of this event was performed and concluded that the HPCI system would have been operable with this condition. If HPCI turbine actuated with the estimated amount of condensate accumulated in the casing and connecting piping, it would have performed its safety function; the HPCI Turbine Exhaust Rupture Disc would not have been challenged by calculated peak pressures; and calculated water hammer loads were within specified load capacities of the turbine flange, downstream piping, struts, snubber, and spring hanger. Based on this, the condition reported in EN 55871 is being retracted. Notified R1DO (Bickett) |
ENS 55593 | 19 November 2021 00:50:00 | FitzPatrick | NRC Region 1 | GE-4 | On November 18, 2021, during the performance of High Pressure Coolant Injection (HPCI) surveillance testing, 23MOV-19 (HPCI PUMP DISCH TO REACTOR INBD ISOL VALVE) did not go open as expected while performing the sensed low water level portion of the test. The ability to manually open 23MOV-19 from the control room was unaffected as such, the HPCI system remained available for use. Failure of 23MOV-19 to open automatically prevents the HPCI system from performing its safety function as such this condition renders HPCI inoperable but available and 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). HPCI inoperable placed the licensee in a 14-day limiting condition for operation for Tech Spec 3.5.1.c. The NRC Resident Inspector was notified. |
ENS 55427 | 24 August 2021 16:51:00 | FitzPatrick | NRC Region 1 | GE-4 | During an extent of condition review of DC control circuits, it was identified there are additional unprotected DC control circuits which are routed between separate Appendix R fire areas. A postulated fire in one area can cause a short circuit and potentially result in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures degrade the degree of separation for redundant safe shutdown trains and are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). Compensatory actions for affected fire areas have been implemented. Design modifications in the affected control circuits are being developed and will be scheduled to correct this condition. |
ENS 54913 | 25 September 2020 15:28:00 | FitzPatrick | NRC Region 1 | A non-licensed contract supervisor had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. | |
ENS 54657 | 10 April 2020 10:58:00 | FitzPatrick | NRC Region 1 | On April 10, 2020, at 0300 (EDT), an oil leak from 23PCV-12, HPCI (High Pressure Core Injection) Trip System Pressure Control Valve (PCV), resulted in the system being declared inoperable. 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 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. | |
ENS 54533 | 20 February 2020 15:04:00 | FitzPatrick | NRC Region 1 | (On February 20, 2020, at 1240 EST, the Licensee determined the following information:) This notification is in reference to reports EN 54130 and LER 2019-002, which were retracted. James A. FitzPatrick Nuclear Power Plant received additional information on the technical basis for the retraction. Further review, including testing of the terminal blocks, demonstrated that the short circuit current would result in heat levels in excess of cable insulation ratings. Unprotected DC control circuits for non-safety related DC motors are routed between separate fire areas. A postulated fire in one area can cause a short circuit and potentially result in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures degrade the degree of separation for redundant safe shutdown trains and are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). Compensatory actions per the Technical Requirements Manual (TRM) for affected fire areas have been implemented. A modification to install fuses in the control circuits for 94P-2(M), 31P-7A(M), 31P-7B(M), and 94P-13(M) has been scheduled and shall correct this condition. The NRC Resident Inspector has been notified. | |
ENS 54503 | 31 January 2020 09:46:00 | FitzPatrick | NRC Region 1 | At 0555 (EST), on January 31, 2020, James A. FitzPatrick was at 38 percent power when an automatic scram occurred as a result of a main turbine trip on high Reactor Pressure Vessel (RPV) water level. The plant was at reduced power in preparation for maintenance activities. The 'A' Reactor Feed Pump (RFP) was being removed from service when a perturbation in reactor water level reached the high RPV water level setpoint. This resulted in a main turbine trip and 'B' RFP trip. The automatic scram inserted all control rods. A subsequent low water level resulted in a successful Group 2 isolation. The plant is stable in Mode 3 with the 'B' RFP maintaining RPV water level. The initiation of the reactor protection systems (RPS) due to the automatic scram signal 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). The licensee notified the NRC Resident Inspector, and the State and Local government for the scram. Decay heat is being removed via the main condenser. | |
ENS 54130 | 24 June 2019 21:18:00 | FitzPatrick | NRC Region 1 | EN Revision Text: POTENTIAL UNANALYZED CONDITION DUE TO UNPROTECTED CONTROL CIRCUITS RUNNING THROUGH MUTILPLE FIRE AREAS During a review of industry Operating Experience it was identified that there were unprotected DC control circuits for non safety-related DC motors which are routed from the Battery Charger Rooms to other separate fire areas. Circuit Breakers used to protect the motor power conductors appear to be inadequate to protect the control conductors. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one area can cause short circuits potentially resulting in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable as an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Requirements of the Technical Requirements Manual (TRM) for the affected fire areas will be implemented." The licensee has notified the NRC Resident Inspector.
In accordance with NUREG-1022, Sections 2.8 and 5.1.2, James A. FitzPatrick Nuclear Power Plant is retracting (formally withdrawing) Licensee Event Report (LER) Number 2019-002. LER 2019-002 was transmitted to the NRC via letter JAFP-19-0080 dated August 23, 2019. The LER reported, under 10 CFR 50.73(a)(2)(ii)(B), the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. Subsequent to submittal of LER 2019-002, FitzPatrick Engineering completed analyses using more accurate input conditions. This analysis has determined no credible hot short scenario will result in damage to adjacent cables in other fire zones, showing that the postulated condition would not degrade plant safety. Therefore, James A. FitzPatrick Nuclear Power Plant is retracting LER 2019-002 (and this event notification). The licensee will notify the NRC Resident Inspector and the New York State Public Service Commission. Notified R1DO (DeFrancisco). | |
ENS 53828 | 16 January 2019 08:12:00 | FitzPatrick | NRC Region 1 | On 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 53778 | 5 December 2018 17:06:00 | FitzPatrick | NRC Region 1 | At 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 52664 | 4 April 2017 11:32:00 | FitzPatrick | NRC Region 1 | GE-4 | On 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 52645 | 28 March 2017 14:47:00 | FitzPatrick | NRC Region 1 | GE-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.
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 52503 | 22 January 2017 16:19:00 | FitzPatrick | NRC Region 1 | GE-4 | Information 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 52490 | 14 January 2017 13:09:00 | FitzPatrick | NRC Region 1 | GE-4 | At 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 52480 | 10 January 2017 22:45:00 | FitzPatrick | NRC Region 1 | GE-4 | This 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.
The licensee is retracting this report based on the following: The testing vendor Argo Turboserve Corp provided input to this condition in Non Conformance Report NC161. Engineering reviewed NC161 and discussed this condition with the vendor. The power supplies showed an initial drift following a 100 hour burn-in but continued operation was stable. Therefore, even though the initial voltage drift did not meet the power supply's calibration requirement in F180-0309, it would not be expected to drift significantly further. Engineering concluded that this condition will have a negligible impact on system components. Readings were steady and within the tolerance of the RHR instrument specification. Based on all current information, the component will function over its mission time of 100 days.Based on these results, a notification for failure to comply or defect per 10 CFR 21.21(d)(3)(i) is not required and the interim report per 10 CFR 21.21(a)(2) may be retracted." The licensee informed the NRC Resident Inspector. Notified R1DO (Bickett) and Part 21 Group via email. |
ENS 52343 | 2 November 2016 15:35:00 | FitzPatrick | NRC Region 1 | GE-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.
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 52329 | 28 October 2016 17:36:00 | FitzPatrick | NRC Region 1 | GE-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.
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 52122 | 25 July 2016 13:43:00 | FitzPatrick | NRC Region 1 | GE-4 | A 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 52073 | 8 July 2016 08:48:00 | FitzPatrick | NRC Region 1 | GE-4 | Oil 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 52045 | 26 June 2016 23:08:00 | FitzPatrick | NRC Region 1 | GE-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.
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 52042 | 24 June 2016 16:06:00 | FitzPatrick | NRC Region 1 | GE-4 | At 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 51985 | 7 June 2016 13:07:00 | FitzPatrick | NRC Region 1 | GE-4 | At 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 51722 | 10 February 2016 05:42:00 | FitzPatrick | NRC Region 1 | GE-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 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 51694 | 29 January 2016 11:02:00 | FitzPatrick | NRC Region 1 | GE-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.
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 51680 | 24 January 2016 01:10:00 | FitzPatrick | NRC Region 1 | GE-4 | At 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 51613 | 19 December 2015 00:44:00 | FitzPatrick | NRC Region 1 | GE-4 | On 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 51595 | 9 December 2015 22:28:00 | FitzPatrick | NRC Region 1 | GE-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.
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 51579 | 2 December 2015 14:14:00 | FitzPatrick | NRC Region 1 | GE-4 | On 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 51512 | 3 November 2015 16:19:00 | FitzPatrick | NRC Region 1 | GE-4 | On 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 51480 | 19 October 2015 08:19:00 | FitzPatrick | NRC Region 1 | GE-4 | On 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 51409 | 18 September 2015 18:01:00 | FitzPatrick | NRC Region 1 | GE-4 | At 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 51405 | 17 September 2015 16:24:00 | FitzPatrick | NRC Region 1 | GE-4 | At 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 51242 | 20 July 2015 14:27:00 | FitzPatrick | NRC Region 1 | GE-4 | On 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 51151 | 12 June 2015 14:05:00 | FitzPatrick | NRC Region 1 | GE-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.
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 51150 | 11 June 2015 23:37:00 | FitzPatrick | NRC Region 1 | GE-4 | At 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 50979 | 12 April 2015 20:51:00 | FitzPatrick | NRC Region 1 | GE-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.
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 50579 | 30 October 2014 13:05:00 | FitzPatrick | NRC Region 1 | GE-4 | On 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 50558 | 22 October 2014 14:05:00 | FitzPatrick | NRC Region 1 | GE-4 | A 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. |