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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.

• * * RETRACTION ON 06/20/2024 AT 1423 EDT FROM CAMERON KELLER TO ROBERT THOMPSON * * *

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).

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.

• * * UPDATE ON 10/4/22 AT 2047 EDT FROM ANDREW WEAVER TO KERBY SCALES * * *

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).

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.

• * * RETRACTION FROM ROBERT GRAHAM TO HOWIE CROUCH AT 2045 EDT ON 9/30/19 * * *

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).

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).

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).

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.

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).

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).

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).

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)