Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 50509 | 4 October 2014 22:02:00 | 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.
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 49551 | 18 November 2013 12:08:00 | James 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 49491 | 31 October 2013 20:41:00 | Per 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 48907 | 10 April 2013 11:06:00 | 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.
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 48654 | 7 January 2013 14:59:00 | 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.
The TSC ventilation system was returned to service at 1650 EST. The NRC Resident Inspector has been notified. R1DO (Newport) notified. |
| ENS 48479 | 5 November 2012 00:40:00 | The reactor was scrammed on a valid reactor protection system activation caused by a main turbine trip. The cause of the main turbine trip is under investigation. All control rods fully inserted. All isolations and initiations occurred as designed. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated as expected. RCIC injected into the reactor coolant system, HPCI did not, as expected. This scram was characterized as uncomplicated and the reactor is stable in Mode 3. The plant is in a normal post shutdown electrical lineup. All systems functioned as required. The NRC Resident Inspector has been notified. |
| ENS 47128 | 5 August 2011 21:14:00 | The 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 one Oswego County Emergency notification siren at approximately 1932 EDT on 8/5/2011. Initial notification to the James A Fitzpatrick control room of the siren activation was via the RECS (Radiation Emergency Communication System) Notification System. Power was removed to the affected siren by off-site repair personnel at 1948 EDT. The affected siren provides coverage to Oswego County. The sirens are utility owned and shared with the Nine Mile Point site. In the event the sirens are needed, the county has its Hyper-Reach (911 call back system) on standby. This issue has been entered into the station corrective action program. The NRC Sr. Resident Inspector has been notified. The licensee notified the State of New York and Oswego County. See similar EN #47127. |
| ENS 46822 | 5 May 2011 02:06:00 | A planned maintenance evolution at the James A. FitzPatrick (JAF) Nuclear Power Plant will remove the Technical Support Center (TSC) ventilation system from service. The TSC ventilation system will be rendered non-functional during the course of the work activity. The TSC ventilation is expected to be out of service for approximately 9 hours from 0400 to 1300 today. 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 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 one the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified.
As of 1215 EDT on 05/05/11, the planned maintenance activity on the TSC ventilation system is complete. Post work testing is complete and the system has been restored to its fully functional state. The NRC Resident Inspector has been informed. Notified R1DO (Harold Gray) |
| ENS 46613 | 14 February 2011 03:09:00 | A planned maintenance evolution at the James A. FitzPatrlck (JAF) Nuclear Power Plant will remove the Technical Support Center (TSC) ventilation system from service. Therefore, the TSC ventilation system will be rendered non-functional during the course of the work activities. The TSC ventilation is expected to be out of service for approximately 13 hours from 0700 to 2000 today. 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 10 CFR 50.72 (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.
As of 1839 hours today, the planned maintenance activity on the Technical Support Center (TSC) ventilation system is complete. Post-work testing is complete and the system has been restored to fully functional. The NRC Resident Inspector has been notified. Notified R1DO (Ferdas). |
| ENS 46355 | 23 October 2010 02:54:00 | The High Pressure Coolant Injection (HPCI) system was declared inoperable due to an instrument power supply failure. The cause of the failure is under investigation. All other ECCS, Emergency Diesels, and Reactor Core Isolation Cooling (RCIC) are operable. The licensee has notified the NRC Resident Inspector. |
| ENS 43752 | 28 October 2007 03:24:00 | High winds (in excess of 40 MPH) resulted in significant debris at the plant intake. Traveling screens had been placed in 'Fast' speed and continuous wash on the previous shift in anticipation of predicted high wind conditions. As the debris entered the intake, high traveling screen differential pressure and lowering intake level prompted the operating crew to enter the appropriate Abnormal Operating Procedure (AOP). Efforts to manually clean the screens were ineffective, and at 0059 hours, the crew inserted a manual scram as directed by procedure at 240 foot intake level. On scram, as expected, reactor vessel level lowered to the low level scram setpoint (177 inches above top of active fuel). At this point, as expected, an automatic Reactor Protection System (RPS) actuation, and a Group 2 Primary Containment Isolation System (PCIS) isolation occurred with no anomalies noted. During plant cooldown reactor level lowered to 177 inches above TAF. This resulted in a valid RPS actuation and a PCIS Group 2 isolation signal. All systems responded as expected. Operators were able to maintain reactor vessel level above the actuation setpoint for High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems, and these systems were not required to operate. The cause of the failure of the traveling screens to maintain intake level is under investigation. All control rods are fully inserted and the plant is stable in Mode 3, Hot Shutdown. Decay heat is being removed via the turbine bypass valves to the main condenser. No SRVs lifted during the transient. The plant is in a normal shutdown electric plant lineup. The licensee notified the NRC Resident Inspector and the New York Public Service Commission. |
| ENS 40684 | 18 April 2004 23:45:00 | At 1830 hours on 18 April 2004, Nine Mile Point Unit 2 was restoring from reactor pressure vessel leakage test. Reactor pressure was ~82 psi, temperature ~178 degrees, reactor level was solid with both reactor recirculation pumps running in slow speed per the leak test procedure. 'B' residual heat removal system was being warmed up in preparation for going in to service. Reactor water cleanup reject from the vessel was secured to maintain RPV (Reactor Pressure Vessel) pressure stable in order to provide driving head for flow through the residual heat removal discharge line to radwaste. The line is warmed up from the reactor, back through the Shutdown cooling isolation valve, to radwaste prior to placing shutdown cooling in service. When warm-up criteria are met, the operating procedure directs securing flow. When flow was secured, this effectively isolated the solid reactor vessel, resulting in a rise in reactor pressure. Pressure peaked at ~146 psi before operators established reactor water cleanup reject flow. When RPV pressure reached 128 psi, the residual heat removal system isolation was automatically initiated as designed. The shutdown cooling injection valve, which was open to support piping warm-up, closed as designed. All other shutdown cooling valves were closed prior to the event per the warm-up lineup. RPV is currently depressurized with shutdown cooling in service. All systems are functioning as expected. The licensee notified the NRC Resident Inspector.
The following is a retraction of ENS Notification #40684: On April 18, 2004, Nine Mile Point Unit 2 reported Residual Heat Removal Shutdown Cooling System isolated due to a pressure spike while warming up the piping. The shutdown cooling injection valve, which was open to support piping warm-up, closed as designed. All other shutdown cooling valves were closed prior to the event per the warm-up lineup. Accordingly, Control Room personnel conservatively initiated ENS reporting under 10CFR50.72(b)(3)(iv)(A) in response to the valid containment isolation signal. Subsequent analysis has concluded that, although the isolation signal was valid, it was not a general containment isolation signal (i.e., the isolation signal only involved the shutdown cooling system) and the shutdown cooling function of the Residual Heat Removal System is not listed under 10CFR50.72(b)(3)(iv)(B) as a system that is required to be reportable. This system isolation does not meet the reporting requirements of 10CFR50.72(b)(3)(iv)(A). Therefore, this notification made on April 18, 2004, is being retracted. The NRC Resident Inspector was notified. NRC R1DO (Ron Bellamy) notified. |
| ENS 40487 | 29 January 2004 13:50:00 | While attempting to perform the Control Room Special Filter Train Functional Test for Division ll, the required train flow test conditions could not be established. Investigation revealed that the inlet backdraft damper 2HVC*DMPV93 was latched closed and would not open automatically on a Special Filter Train start. The damper was found latched closed due to an improperly set spring-loaded 'stop.' As a result, the Division II Control Room Special Filter Train was declared inoperable as of 1530 on 1/28/04. Technical Specification 3.7.2, Condition A was entered which allows 7 days to restore the Division II Control Room Special Filter Train to an operable status. A review of plant computer data and monthly surveillance tests for the Division II Special Filter Train determined that the backdraft damper has been latched closed since October 2003. As a result, the Division Il Control Room Special Filter exceeded Technical Specification 3.7.2 action to restore Control Room Special Filter Train Subsystem to operable within 7 days. This is reportable under 50.73(a)(2)(i)(B) 'Any operation or condition which was prohibited by the plant's Technical Specifications,' and Section 2.F of the Nine Mile Point Operating License. Per Nine Mile Point Operating license, initial notification that the facility is not operated in accordance with the Technical Specifications shall be made within 24 hours to the NRC Operations Center via the Emergency Notification System, with written follow up within 30 days in accordance with 10 CFR 50.73(b), (c), and (e). A repair plan is being prepared with work to commence today (1/29). Following repairs, surveillance testing will be performed to confirm proper operation of the damper. An inspection of the redundant Division I Control Room Special Filter Train damper and review of monthly surveillance tests demonstrated that the latching mechanism does not interfere with the ability of the damper to open as designed. Therefore, operability of the Division I Control Room Special Filter Train has been confirmed. The licensee informed the NRC Resident Inspector. |