Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 56501 | 2 May 2023 19:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Communications | The following information was provided by the licensee via email: At approximately 1500 (EDT) on 5/2/2023, it was determined that the commercial telecommunications capacity was lost to the Palisades Nuclear Plant (PNP) control room and technical support center due to an issue with the telecommunications provider. After discovery of the condition it was discovered that this loss also included the emergency notification system (ENS). Communications link via the satellite phone was tested satisfactorly. In addition, if needed, the satellite phone would be used to initiate call-out of the emergency response organization. The condition did not affect the ENS or commercial telecommunications capabilities at the offsite Emergency Operations Facility. The telecommunications provider has not provided an estimated repair time. PNP will be notifying the NRC resident inspector.
The following information was provided by the licensee via email: This notification is being made to retract event EN 56501 that was reported on May 02, 2023. Based on further investigation, the Emergency Plan and Emergency Implementing Procedures provide an acceptable alternative routine communication system, which is satellite phones, for communicating with Federal, State, and local offsite agencies, that are in addition to the primary commercial telephone system. It was determined that no actual or potential loss of offsite communications capability existed per 10 CFR 50.72(b)(3)(xiii). This is consistent with NUREG 1022, Revision 3, Supplement 1, 'Event Report Guidelines 10 CFR 50.72(b)(3)(xiii),' and NEI 13-01, Revision 0, 'Reportable Action Levels for Loss of Emergency Preparedness Capabilities.' The NRC Decommissioning Inspector has been notified of the retraction. Commercial telecommunications to the plant were restored at approximately 0600 EDT on 5/3/2023. Notified R3DO (Orlikowski) | |
| ENS 55311 | 16 June 2021 19:50:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Atmospheric Steam Dump Valves Inoperable | On June 16, 2021, at 1550 EDT, Palisades Nuclear Plant was operating in Mode 1 at 100% power. At that time, operations identified an acrid odor in the control room. Investigation revealed that the steam dump control relay had failed, rendering all four atmospheric steam dump valves inoperable. The loss of function of all four atmospheric steam dump valves is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D) as an 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 mitigate the consequences of an accident. Troubleshooting and replacement of the relay are in progress. The plant remains stable in Mode 1 at 100% power. The NRC Resident Inspector has been notified. Unit 1 is in a 24 hour LCO for Tech Spec 3.7.4.b, atmospheric steam dump valve inoperability. The Unit is in a normal offsite power line-up. | |
| ENS 54897 | 11 September 2020 23:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition | At 1930 EDT, on September 11, 2020, Palisades Nuclear Plant was conducting ultrasonic data analysis from reactor vessel closure head in-service inspections. During this analysis, signals that display characteristics consistent with primary water stress corrosion cracking were identified in head penetration 34. No leak path signal was identified during ultrasonic testing. The plant was in cold shutdown at 0% power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. This is the only indication that is currently present, however, if additional indications are found, they will also be repaired prior to the plant startup. The licensee notified the NRC Senior Resident Inspector. | |
| ENS 54867 | 31 August 2020 00:40:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty (Ffd) Policy Violation | A contract employee supervisor had a confirmed positive for alcohol during a fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. | |
| ENS 53819 | 9 January 2019 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip from Full Power Due to Rps Testing | At 1034 EST on January 9, 2019, with the reactor at 100% power, an automatic reactor trip was initiated. The trip occurred while Reactor Protection System testing was in progress. The trip was uncomplicated with all systems responding normally following the rip. Troubleshooting and investigation of the cause is ongoing. All full-length control rods inserted fully. Auxiliary Feedwater System actuated as designed in response to low steam generator water levels. Operations stabilized the plant in Mode 3 (hot standby). Decay heat is being removed by the turbine bypass valve. This condition has no impact to the health and safety of the public. The licensee notified the NRC Resident Inspector. | |
| ENS 53813 | 3 January 2019 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Cycling of Turbine Governor Valve | At 2028 (EST) on January 3, 2019, with the reactor at 85% power, the reactor was manually tripped due to cycling of Turbine Governor Valve #4. The trip was uncomplicated with all systems responding normally following the trip. Investigation of the cause of the valve cycling is ongoing. All full-length control rods inserted fully. Auxiliary Feedwater System actuated as designed in response to low steam generator water levels. Operations stabilized the plant in Mode 3 (hot standby). Decay heat is being removed by atmospheric dump valves. This condition has no impact to the health and safety of the public. The licensee notified the NRC Senior Resident Inspector. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). | |
| ENS 53749 | 21 November 2018 05:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Identified During Surface Inspection of Reactor Head Nozzle Penetration | On November 21, 2018, during an extent of condition review, after completion of ultrasonic testing, further interrogation of reactor vessel closure head (RVCH) penetration 36 was performed using eddy current testing. The testing detected three repairable indications. No indication of boric acid leakage was identified at this location during the bare metal visual inspection. Extent of condition review is complete on all RVCH penetrations. The plant was in cold shutdown at 0 percent power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector. | |
| ENS 53734 | 11 November 2018 04:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Indication During Ultrasonic Inspection of Reactor Head Nozzle Penetration | On November 11, 2018, during ultrasonic data analysis from reactor vessel closure head in-service inspections, signals that display characteristics consistent with primary water stress corrosion cracking in head penetration 33 were identified. No indications of boric acid leakage and no surface indications were detected at this location during bare metal visual inspection. The plant was in cold shutdown at 0% power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector. | |
| ENS 53733 | 10 November 2018 05:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Boric Acid Identified on Reactor Vessel Head Penetration | On November 10, 2018, during a planned bare metal visual inspection of the reactor head, boric acid was discovered at a CRDM (Control Rod Drive Mechanism) nozzle to reactor head penetration. Investigation of the source of the boric acid is ongoing. The plant was in cold shutdown at 0% power and Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. All other reactor vessel head penetrations have had a bare metal visual inspection completed with no other indications identified. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector. | |
| ENS 52764 | 19 May 2017 06:06:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Reactor Protection System Actuation While Shutdown | Pursuant to 10CFR50.72(b)(3)(iv)(A), notifications are being performed for a valid actuation of the reactor protection system resulting in a full scram. The actuation was a result of pre-startup testing. The generator coastdown protective relay was left in service which needed to be bypassed to facilitate the testing. This resulted in a reactor scram occurring. The reactor was subcritical with all rods inserted at the time of the actuation. All systems functioned as designed. The licensee notified the NRC Resident Inspector. | |
| ENS 52722 | 2 May 2017 13:28:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Failed Ultrasonic Testing of Weld | On May 2, 2017, during planned inspections, an ultrasonic examination performed on weld PCS-4-PRS-1P1-1, revealed an axial indication in the pressurizer nozzle to safe end area of the weld. This indication does not meet applicable acceptance criteria under ASME, Section XI. The plant was in cold shutdown at 0% power for a planned refueling outage at the time of discovery. The condition will be resolved prior to plant startup. This condition has no impact to the health and safety of the public. The licensee notified the NRC Senior Resident Inspector. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A), since an indication was found that did not meet acceptance criteria referenced in ASME Code, Section XI.
Additional evaluations of the recorded indication concluded that the indication was attributed to an erroneous ultrasonic response. This was the result of a combined effect of compromised surface contact at the area of the recorded indication and associated examination scan speed. The contact issue is attributed to the specific tooling configuration required for this exam. The combination of these factors resulted in the introduction of an erroneous reflector in the area of interest that had characteristics of a relevant indication. The vendor repeated the entire examination for axial flaws and there were no service induced indications recorded. A review of the newly acquired data by site, vendor and EPRI personnel confirmed that no service induced flaws are present. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Hills). | |
| ENS 52647 | 29 March 2017 19:08:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Discovery of Non-Conforming Conditions During Tornado Hazards Analysis | During an evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Palisades Nuclear Plant personnel identified conditions in the plant design such that specific TS equipment is considered not adequately protected from tornado missiles. Specifically, vulnerabilities were identified in the following systems and components: Service Water System - Service water pump discharge header and service water pump cable trays. Fuel Oil Transfer System - Fuel oil transfer piping and transfer pump cable trays. Emergency Diesel Generators - Vent lines on the fuel oil day tanks. Control Room Heating, Ventilation, and Cooling System - Both the normal and emergency intake ducts. Steam Driven Auxiliary Feedwater Pump - Feedwater pump relief valves. Component Cooling Water System - Component cooling water surge tank. The identified vulnerabilities are being addressed in accordance with Enforcement Guidance Memorandum (EGM), 15-002, and Interim Staff Guidance, DSS-ISG-2016-01. Initial compensatory measures are in place. The licensee notified the NRC Resident Inspector. | |
| ENS 51922 | 12 May 2016 11:55:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness-For-Duty Policy Violation | A non-licensed contractor supervisor violated the fitness-for-duty policy (for alcohol) during a random fitness-for-duty test. The contractor's access to the plant has been terminated. The NRC Resident Inspector has been notified. | |
| ENS 51820 | 24 March 2016 06:11:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Ventilation Declared Inoperable | At approximately 0211 (EDT), on March 24, 2016, both control room ventilation filtration trains were declared inoperable in accordance with Technical Specification 3.7.10, Condition B, due to a control room boundary door not being fully closed. Following routine security rounds, the door was unable to be fully closed due to the door's locking bolts not retracting back into the door body, causing interference between the door and door frame. Mitigating actions have been implemented that ensure control room envelope (CRE) occupant radiological exposures will not exceed limits, and CRE occupants are protected from chemical and smoke hazards. Repairs to the door are currently in progress. Technical Specification 3.7.10 allows control room boundary doors to be opened intermittently, under administrative control for preplanned activities, provided the doors can be rapidly restored to the design condition. Previous evaluations of the door not being fully closed for a limited time concluded no loss of safety function had existed. This condition had no impact on the health and safety of the public. The NRC Resident Inspector has been notified. | |
| ENS 51397 | 16 September 2015 05:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | At 0117 (EDT) on 9/16/2015 a reactor trip occurred (4-hr non-emergency). The plant was at approximately 85% power performing a coastdown in preparation for a refueling outage when a Digital Electro-Hydraulic (DEH) alarm was received in the control room. Shortly following receipt of the alarm the turbine tripped. This resulted in an RPS actuation and a reactor trip on Loss of Load. The crew entered EOP-1 Standard Post Trip Actions and completed all required actions. The crew subsequently entered EOP-2 Reactor Trip Recovery. All full-length control rods inserted fully. Auxiliary Feedwater System actuated in response to low steam generator water levels (8-hr non-emergency). Steam generator water levels are in progress of being returned to normal operating levels. No known primary to secondary leakage. Atmospheric Steam Dump Valves lifted after the trip and subsequently reseated. The plant is currently stable in Mode 3 at NOP/NOT being maintained by the Turbine Bypass Valve. Initial investigation into the cause of the turbine trip appears to be from a DEH power supply failure. The NRC Resident Inspector was notified of the reactor trip at 0139 on 9/16/2015. | |
| ENS 51224 | 13 July 2015 12:38:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Spill of Portable Chemical Toilet to the Environment | On 7/13/15 at 1000 EDT, a spill to the environment was determined to be reportable to the state environmental and local health agencies. The spill occurred when a portable chemical toilet tipped over and was identified at approximately 0838 EDT. The contents and exact quantity of the spill are unknown, but the toilet has a capacity of 60 gallons. The discharge flowed to a storm drain which ultimately discharges to the beach of Lake Michigan. Rainfall was present when the spill was identified. Cleanup efforts are in progress. The State of Michigan, via the Pollution Emergency Alert System (PEAS), was notified as required by the Site Spill Plan by the Site Environmental Coordinator at 1024 (EDT). The Local government (Van Buren County) was notified at 1045 (EDT) via 911. The NRC Senior Resident Inspector has been notified. | |
| ENS 51033 | 2 May 2015 16:23:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Seismic Activity Felt on Site | At 1241 EDT, Operations staff at Palisades declared an Unusual Event under EAL HU1.1 due to seismic activity felt on site. No seismic alarms were initiated. No plant equipment was affected. The epicenter of the 4.2 magnitude earthquake was located south of Galesburg, MI. Palisades continues to operate at 100% power. The licensee has notified the NRC Resident Inspector.
The licensee terminated the Unusual Event at 1541 EDT on 5/2/15. The licensee has notified the NRC Resident Inspector and the state and local government. Notified R3DO (Orlikowski), IRD MOC (Stapleton), NRR EO (Morris), NRR ET (Dean), and R3RA (Pederson). Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA). | |
| ENS 50907 | 19 March 2015 20:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification of Tritium in Ground Water | On February 26, 2015, as a result of routine monitoring well sampling, Palisades Nuclear Plant identified tritium in temporary wells 7 and 8, located within the plant protected area. Tritium concentrations were less than the threshold value (20,000 pCi/L) for initiating voluntary communications in accordance with Nuclear Energy Institute Ground Water Protection Initiative. The station promptly isolated and rerouted the likely source. Subsequent sample concentrations from samples obtained on March 18, 2015, have resulted in concentrations of tritium less than the minimum detectable activity. The wells are currently used only for on-site sampling and not for drinking water. There is no threat to public health and safety. These results confirm that a leak from the pipe that runs from the turbine sump oil separator to the turbine building drain tank was the likely cause. The volume of the leak cannot be determined but is potentially greater than 100 gallons. Therefore, voluntary communications have been made to state and local stakeholders. The Licensee has notified the Michigan Department of Environmental Quality, Van Buren County Administrator, the Township Supervisors in Covert, Geneva, and South Haven Townships, as well as the City of South Haven Mayor and City Manager. The licensee has notified the NRC Resident Inspectors. The tritium is suspected to be migrating through the steam generator tubes to the turbine building waste water. | |
| ENS 50100 | 8 May 2014 22:24:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Made Due to Oil Leak | At 1824 hours (EDT) on May 8, 2014, an oil leak was discovered on EX-07 Safeguards Transformer 1-1 that sprayed outside the transformer containment berm. The oil released to the ground was approximately 70 gallons, which is greater than the reporting requirements. The oil was not released to any surface water and has no drain path to reach any surface water. The State of Michigan, via the Pollution Emergency Alert System (PEAS) and Van Buren County was notified as required by the Site Spill Plan by the Control Room at 2003 hours. The licensee has notified the NRC Resident Inspector. | |
| ENS 49924 | 17 March 2014 18:18:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Displayed Meteorological Data Potentially Inaccurate | At 1418 EDT on 03/17/2014, it was determined the atmospheric stability classification data display on the plant process computer (PPC) could have been potentially inaccurate. In September 2012, implementation of a hardware and software modification to separate the meteorological tower computer from the PPC introduced a calculation error to the PPC software design that resulted in the PPC potentially displaying inaccurate stability class data. The ability to accurately perform dose assessment calculations could potentially be affected when using the stability class indication obtained from the PPC. The availability and accuracy of backup stability class data via the meteorological tower computer or from Weather Services International were not affected. Within 24 hours of discovery, the procedurally described alternatives for obtaining stability class data were implemented. Subsequently, the PPC stability class software calculation was corrected and the PPC stability class display restored for use. There was no impact to any emergency declaration because there were no actual emergencies from the time the modification was implemented until the error was corrected. The licensee has notified the NRC Resident Inspector. | |
| ENS 49960 | 8 February 2014 04:19:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of a Containment Isolation Signal | This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(2)(iv)(A) to provide information pertaining to the invalid actuation of a containment isolation signal. On February 8, 2014, at 0019 (EST), with the plant in Mode 6 and fuel moves in-progress, an invalid containment isolation actuation signal was generated that affected containment isolation valves in more than one system and caused both trains of the control room heating, ventilation and cooling systems to swap from normal to emergency mode of operation. For both occurrences, all necessary follow-up actions were taken in accordance with the abnormal operating procedure for a spurious containment isolation event. All equipment responded in accordance with the plant design. The invalid actuation signal was caused by spurious upscale spikes in fuel handling area radiation monitor, RIA-2316. Actual radiation levels in the vicinity of RIA-2316 were verified to be normal and below the alarm set point. The indication on RIA-2316 spiked momentarily above the high alarm set point and then returned to normal levels. No alarms were received from the redundant fuel handling area radiation monitor, RIA-2317. The NRC Resident Inspector has been notified. | |
| ENS 49959 | 6 February 2014 15:43:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of a Containment Isolation Signal | This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(2)(iv)(A) to provide information pertaining to the invalid actuation of a containment isolation signal. On February 6, 2014, at 1143 (EST), with the plant in Mode 6 and fuel moves in-progress, an invalid containment isolation actuation signal was generated that affected containment isolation valves in more than one system and caused both trains of the control room heating, ventilation and cooling systems to swap from normal to emergency mode of operation. For both occurrences, all necessary follow-up actions were taken in accordance with the abnormal operating procedure for a spurious containment isolation event. All equipment responded in accordance with the plant design. The invalid actuation signal was caused by spurious upscale spikes in fuel handling area radiation monitor, RIA-2316. Actual radiation levels in the vicinity of RIA-2316 were verified to be normal and below the alarm set point. The indication on RIA-2316 spiked momentarily above the high alarm set point and then returned to normal levels. No alarms were received from the redundant fuel handling area radiation monitor, RIA-2317. The NRC Resident Inspector has been notified. | |
| ENS 49796 | 5 February 2014 18:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Discovered on Pressurizer Nozzle Weld During Testing | On February 5, 2014, during planned inspections, an ultrasonic examination performed on weld PCS-6-PRS-1C1-1 (RV-1041) revealed two axial indications in the root area of the weld. The weld containing the indications is the nozzle to safe end dissimilar metal weld on the flange for pressurizer safety valve RV-1041. These two indications do not meet applicable acceptance criteria under ASME, Section XI, IWB-3600, 'Analytical Evaluation of Flaws,' or ASME Section Xl, Table IWB-3410, 'Acceptance Standards,' and will require a repair or replacement activity in order return the weld to an acceptable condition. The plant was in cold shutdown at 0% power for a planned refueling outage at the time of discovery. Replacement or repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. The licensee notified the NRC Senior Resident Inspector. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A), since indications were found that did not meet acceptance criteria referenced in ASME Code, Section XI. | |
| ENS 49773 | 29 January 2014 14:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Indications Identified on Control Rod Drive Mechanism Housings | On January 29, 2014, during planned inspections of control rod drive mechanism (CRDM) upper housings, it was determined that the inspection results for some housings did not meet the applicable acceptance criteria. That is, evaluations of the housing indications are being performed under ASME Code, Section XI, IWB-3600, 'Analytical Evaluation of Flaws,' and indications were identified that exceed acceptance criteria specified in the Code. None of the indications were through-wall and there was no evidence of leakage. The housing indications, varying in depth and length characteristics, were identified in 17 of the 45 CRDM housings inspected. All 45 CRDM housings were inspected, which constituted 100% extent of condition inspection. The plant was in cold shutdown at 0% power for a planned refueling outage at the time of discovery. Replacement or repair actions are in progress and will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10CFR 50.72(b)(3)(ii)(A), since indications were found that did not meet acceptance criteria referenced in ASME Code, Section XI. The licensee notified the NRC Senior Resident Inspector. | |
| ENS 49516 | 7 November 2013 20:52:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits has determined the described condition to be applicable to the Palisades Nuclear Plant resulting in a potentially unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1E batteries ampere indications, located in the cable spreading room at Palisades, do not include overcurrent protection features to limit the fault current. In the postulated event, a fire could cause one of the ammeter wires to short to ground. Simultaneously, it is postulated that the fire causes another DC wire from the opposite polarity on the same battery to also short to ground. This could cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (i.e., heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. Interim compensatory measures (i.e., fire tours) have been implemented for affected areas of the plant. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. | |
| ENS 49298 | 22 August 2013 14:48:00 | 10 CFR 26.719, FFD Reporting requirements | Violation of the Fitness for Duty Program | A licensed employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee informed the NRC Resident Inspector and will inform stakeholders at their scheduled meeting. | |
| ENS 49276 | 13 August 2013 15:02:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Control Room Ventilation Filtration Trains Declared Inoperable | At approximately 1102 (EDT), August 13, 2013, both control room ventilation filtration trains were declared inoperable in accordance with Technical Specification 3.7.10, Condition B, due to a control room boundary door not being fully closed. The door was unable to be closed for approximately nine minutes due to an apparent mis-operation of the door operating mechanism. The door's locking bolts fully extended causing interference between the door and door frame. The door was restored to operable status at approximately 1111 (EDT), August 13, 2013. Technical Specification 3.7.10 allows control room boundary doors to be opened intermittently, under administrative control for preplanned activities, provided the doors can be rapidly restored to the design condition. Previous evaluations of the door not being fully closed for a limited time concluded no loss of safety function had existed. This condition had no impact on the health and safety of the public. The NRC Resident Inspector has been notified. | |
| ENS 49002 | 5 May 2013 05:12:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Shutdown - Safety Injection Refueling Water Tank Declared Inoperable Due to Leak | At 0112 EDT on May 5, 2013, the plant commenced a shutdown due to water leakage from the SIRW (Safety Injection Refueling Water) Tank exceeding the operational decision-making issue process trigger point of 34 gallons per day, causing it (the SIRW) to be declared inoperable and requiring entry into Technical Specification LCO 3.5.4 Condition B. LCO 3.5.4 Condition B requires the SIRW Tank to be returned to Operable status within one hour or entry into Condition C which requires the plant to be in Mode 3 within 6 hours and Mode 5 within 36 hours. This event had no impact on the health and / or safety of the public. The NRC Resident has been notified. The exact location of the leakage has not been determined at this time. The Plant will be taken to Mode 5. The licensee has been operating with SIRW leakage at a rate of less than 34 gallons per day. The leakage has increased for unknown reasons to a calculated value of approximately 90 gallons per day. See EN #48018 dated 06/12/12 for similar report on a technical specification shutdown for the SIRW tank leakage. | |
| ENS 48758 | 15 February 2013 18:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Component Cooling Water Train Out of Service | At 2030 hours (EST) on February 14, 2013, technical specification (TS) 3.7.7 condition A was entered due to the right train of the component cooling water (CCW) system being declared inoperable. The cause of the inoperable train was the identification of an approximate 40 gallon per hour CCW system to service water system leak inside the 'A' CCW heat exchanger. TS 3.7.7 condition A requires restoration of the inoperable train within 72 hours. If the restoration is not completed within 72 hours, the plant must be in Mode 3 within 6 hours and in Mode 5 within the subsequent 36 hours. Due to the inability to repair the leak within the required 72 hour time frame during power operation, a plant shutdown was initiated at approximately 1300 hours on February 15, 2013. Entry into Mode 3 is expected at approximately 1700 hours on February 15, 2013. The plant will enter Mode 5 to execute leak repair. Mode 5 entry is expected at approximately 0800 hours on February 16, 2013. This event has no impact on the health and safety of the public. The NRC Resident Inspector has been notified. | |
| ENS 48478 | 4 November 2012 16:30:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Secondary Side Steam Leak | At 1115 EST on November 4, 2012, primary coolant loop #2 was declared inoperable due to a small un-isolable steam leak on a drain valve of an atmospheric steam dump valve on the secondary side of the 'B' Steam Generator. The valve is ASME Class II high energy piping and the non-conforming condition could not be evaluated with the steam generator pressurized. Based on the condition of the valve and the inability to evaluate, Technical Specification 3.4.4, PCS loops - Modes 1 and 2, Required Action A.1 was entered which requires the plant to be placed in Mode 3 in 6 hours. Repair of the valve may require cooldown to Mode 5. At 1230 EST on November, 2012, Palisades initiated a shutdown in accordance with Technical Specification 3.4.4. The licensee has notified the NRC Resident Inspector. | |
| ENS 48182 | 12 August 2012 08:18:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Reported Based on the Discovery of Pressure Boundary Leakage | Following a planned shutdown to investigate the source of elevated Primary Coolant System (PCS) Unidentified Leakage, the Mode 3 PCS walk-down identified a steam leak on CRD-24, Control Rod Drive Mechanism (CRD), pressure housing. The leak is ~ 1 foot above the CRD to Reactor Head flange. The leak was observed from the Refueling floor deck and appears to be coming from an area of the CRD with no bolted connections. Leakage from this area is unexpected and the mechanism of failure is not understood at this time. Closer examination of the leak is expected to occur in parallel with plant cool-down. The plant entered T.S. 3.4.14 (PCS Operational Leakage) Condition B. at 0418 EDT hours this morning and this requires the plant to be placed in Mode 5 within 36 hours. The leakage is considered Pressure Boundary Leakage. Operations is in progress of performing the plant cool-down from Mode 3 to Mode 5. The licensee discovered this condition following a shutdown the morning of 8/12/12 to investigate unidentified primary coolant leakage of about 0.3 gpm that had been recently trending upwards. Plant conditions are currently 400 degrees F with primary system pressure at about 1000 psi. The licensee has notified the NRC Resident Inspect | |
| ENS 48018 | 12 June 2012 18:56:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown | At 1456 hours on June 12, 2012, the plant commenced a shutdown due to water leakage from the SIRW (Safety Injection Refueling Water) tank exceeding the operational decision-making issue process trigger point of 31 gallons per day causing it to be declared inoperable and requiring entry into Technical Specification (TS) 3.5.4, Condition B. TS 3.5.4 Condition B requires the SIRW Tank to be returned to operable status within one hour or entry into Condition C that requires the plant to be in Mode 3 within 6 hours and Mode 5 within 36 hours. Actual leakage from the SIRW Tank was measured at approximately 31.4 gallons per day. This event had no impact on the health and/or safety of the public. The NRC Senior Resident Inspector has been notified. The licensee believes that the tank is leaking from several locations. However, at this time, they cannot determine exact locations. The refueling water has minor tritium contamination. The refueling water is being collected in a reservoir and then pumped into a holding tank. The licensee will be shutting down to cold shutdown.
Subsequent review concluded the condition did not meet reporting criteria. The rate of water leakage from the Safety Injection Refueling Water Tank had reached an administrative limit. Shutdown was not required by the Technical Specifications. The plant shutdown commenced in accordance with normal plant operating procedures. At 1849 hours on June 12, 2012, the reactor was manually tripped. This event is not reportable as the reactor trip was part of a normal shutdown for corrective maintenance. Mode 5 was reached at 1745 hours on June 13, 2012. The NRC Resident Inspector has been notified. Notified R3DO (Lipa). | |
| ENS 47815 | 8 April 2012 09:01:00 | Other Unspec Reqmnt | Offsite Notification of Noise Associated with Plant Shutdown | Notified Van Buren County Sheriff of atmospheric steam dump valve usage for cooldown of Palisades Nuclear Plant for start of refueling outage. The licensee will notify the NRC Resident Inspector. | |
| ENS 47523 | 14 December 2011 20:10:00 | Other Unspec Reqmnt 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Both Main Feedpumps | The reactor was manually tripped at 1510 EST on 12/14/11 due to loss of both main feedpumps. Both feedpumps tripped on low suction pressure due to an apparent unplanned opening of the 'A' main feedpump recirculation valve. The cause of the main feedpump recirculation valve opening has not been determined. All full length control rods fully inserted. Auxiliary feed pump P-8A automatically started at 1511 EST on steam generator level as designed (10CFR50.72(b)(3)(iv)(A)). The turbine bypass valve is in service maintaining reactor coolant system temperature (by directing steam flow to the main condenser). The plant is stable in mode 3 (and the reactor trip was considered uncomplicated). The Van Buren County Sherriff was notified (per other plant requirements) concerning use of the atmospheric steam dump causing excessive noise in the vicinity of the plant (immediately following the plant trip). The plant electric power is in the normal shutdown configuration. There was no primary to secondary leakage. A press release is planned for the local media. The licensee notified the NRC Resident Inspector. | |
| ENS 47322 | 26 September 2011 10:02:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition - Appendix R Equipment May Not Have Been Available | An Appendix R non-compliance issue was identified. When the plant automatically tripped on September 25, 2011, an unexpected automatic trip was identified in DC shunt trip breakers 72-01 and 72-02. This automatic trip function was not considered in the Appendix R electrical coordination evaluation. Preliminary analysis has shown that if the shunt trip breaker would have automatically opened due to fire induced fault currents, then Appendix R credited equipment may have been lost unexpectedly. The operability evaluation performed, resulted in increasing the shunt trip settings to maximum, and implementing compensatory measures to isolate some non-safety related electrical loads, which restored compliance with 10 CFR 50 Appendix R. The licensee notified the NRC Senior Resident Inspector. CR# PLP2011-04835 | |
| ENS 47290 | 25 September 2011 19:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Loss of Two 120 Volt Ac Instrument Busses | At 1506 EDT, while the electricians were working on the left train DC bus, a bus bar slipped causing an arc and a loss of the left train DC busses D-10 L and D-10 R. This resulted in the loss of two preferred AC (120 Volt Instrument) busses Y-10 and Y-30. The loss of both preferred AC busses caused a reactor trip, a safety injection signal, auxiliary feedwater actuation signal, containment high radiation isolation signal, and main steam isolation signal. All systems responded as expected. Electric power has been restored to the affected DC busses and preferred AC busses. The plant is stable in Mode 3 at NOT and NOP, and controlling temperature using Atmospheric Dump Valves. Pressurizer level is high due to the loss of letdown (result of containment isolation signal), however, it is recovering slowly. All rods fully inserted and the electrical lineup is back to normal. The licensee has notified the NRC Resident Inspector, and will be notifying local agencies. The licensee will also be issuing a press release. | |
| ENS 47271 | 16 September 2011 18:50:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Due to Primary System Leakage Greater than 10 Gpm | The Licensee declared an Unusual Event for Palisades Unit 1 on 09/16/2011 at 1450 EDT based on EAL SU 8.1, RCS (Reactor Coolant System) leakage exceeding 10 gallons per minute (gpm). The licensee was monitoring an increase in RCS leakage, and at a rate of 3.5 gpm entered their off normal procedure and began shutting down the plant. Technical Specification requires the plant to be in Mode 3 within 6 hours. Leakage increased to greater than 10 gpm, and at 1454 EDT the reactor was manually tripped from 79% power. All control rods fully inserted, and the shutdown was described by the licensee as uncomplicated. Unit 1 is stable in Mode 3. No safety injection was required since two charging pumps (B&C) were able to keep up with RCS leakage estimated to be between 14 and 15 gpm. Pressurizer level was restored to 43% and rising. RCS pressure was greater than 2000 psi and RCS temperature was being maintained at no load Tave of 535F on the turbine bypass valves. There is no indication of any primary-to-secondary leakage and all equipment is available except for charging pump 'A', which was tagged out of service for planned maintenance. An entry into containment had been made and the licensee had identified the source of the RCS leakage as being in the vicinity of the 'A' pressurizer spray control valve #1057. This was based on a steam plume seen from below the pressurizer looking up through grating towards this valve. The licensee has notified the NRC Resident Inspector. * * * UPDATE FROM JAMES BYRD TO JOHN KNOKE AT 1952 EDT ON 09/16/11 * * * At 1934 EDT the licensee terminated from their Unusual Event due to EAL SU 8.1. The plant is still in Mode 3 with a leak rate of 0.324 gpm..The licensee has confirmed that the leak is a result of the packing gland backing out of pressurizer spray valve #1057. The licensee has notified the NRC Resident Inspector. The R3DO (Bloomer) was notified. Notified FEMA (Eiscoe) and DHS (Flinter). | |
| ENS 47187 | 23 August 2011 18:52:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of an Unusual Event Due to a Seismic Event | The licensee declared a Notification of Unusual Event under EAL HU1.1 due to seismic activity. There were no reports of personnel injury or significant plant damage. The licensee has notified the NRC Resident Inspector and state and local authorities.
The licensee terminated the Notification of Unusual Event at 1825 EDT. The licensee has notified the NRC Resident Inspector and state and local authorities. Notified R3DO (Passehl), IRD (Morris), FEMA (Via) and DHS (Bean). | |
| ENS 46627 | 20 February 2011 00:20:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notifiection Due to Inadvertent Chemical Discharge | At 1920 (EST) on 2-19-11, a chemistry technician notified the Shift Manager that more than the expected amount of Sodium Bisulfite was inadvertently added to the Circulating Water System discharge. The Spill Prevention Control & Countermeasures (SPCC) Plan was referenced for required actions. 17 additional gallons of Sodium Bisulfite was discharged to surface waters of Lake Michigan. This amount of Sodium Bisulfite is less than the reportable quantity of 417 gallons per the SPCC Plan. However, since it was released to surface water (Lake Michigan), a notification to the District Water Quality Division (DWQD) was made at 2020 (EST) by the site's Senior Environmental Specialist (SES). The DWQD requested that additional notifications be made to state (Pollution Emergency Alert System - PEAS) and local (911 operator) agencies. PEAS was notified at 2020 (EST) by the SES and Van Buren County dispatch (911) was notified by the Control Room staff at 2047 (EST). The licensee notified the NRC Resident Inspector. | |
| ENS 46564 | 22 January 2011 22:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Loss of Generator Load | The licensee reported a loss of main generator load at full power resulting in a generator trip, turbine trip, and reactor trip. All rods fully inserted. All safety systems functioned as required. The reactor is stable at no-load temperature and pressure in Hot Standby. Auxiliary feedwater started as expected and is currently supplying cooling water to the steam generators. Decay heat is being removed via the atmospheric steam dumps because the turbine bypass system did not respond as expected. There is no known primary to secondary generator leakage The grid is stable and the plant is in a normal post-trip electrical lineup. The reactor trip was characterized as uncomplicated. The cause of the loss of generator load is not yet know and under investigation. The licensee has notified the NRC Resident Inspector. The local County Sheriff was notified of the use of the atmospheric steam dumps to alleviate any concern from local population in the vicinity of the plant. | |
| ENS 46524 | 8 January 2011 18:03:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Catastrophic Failure of Non-Safety Related Bus Breaker | A notice of unusual event was declared for 'Hazards and other conditions affecting plant safety' at 1303 EST for Emergency Action Level (EAL) HU1 as a result of a breaker/bus fault in a non-safety related feeder bus 'F'. The loss of the bus resulted in loss of cooling water tower pumps and fans and subsequently a loss of one cooling tower. There is indication of a pressure transient on the breaker panel and smoke but no fire. There is no indication of sabotage or terrorism, and no offsite assistance requested. There is no radiological release in progress. The NRC remains in the normal mode. NRC Resident staff are enroute to the site. Plant is stable at 55 % reactor power following a down-power maneuver from 100% reactor power with all safety related equipment operable.
At 0043 Palisades terminated from their Unusual Event for 'Hazards and Other Conditions Affecting Plant Safety'. The faulted bus has been isolated, and licensee started their investigation into the cause of this incident. The licensee has tagged out and racked out the breaker, and has isolated the startup power to the 'F' bus. The associated Startup Transformer has also been isolated. The 'F' bus and associated Startup Transformer will not be energized until the situation is better understood, and repaired if necessary. Power level will be maintained at 53% due to the loss of associated equipment supporting one cooling tower. The licensee is in a 72 hour LCO. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Christine Lipa), IRD MOC (Bill Gott), NRR EO (Mike Cheok), DHS (Gates), FEMA (Casto) | |
| ENS 46505 | 23 December 2010 01:00:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Report Involving a Supervisor | A licensed operator had a confirmed positive for a Controlled Substance during a random fitness-for-duty test. The employee's unescorted access has been suspended. Contact the Headquarters Operations Officer for additional details. The NRC Resident Inspector has been notified. | |
| ENS 46302 | 1 October 2010 17:25:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition - Appendix R Non-Compliance Issue Identified | At 1325 EDT on October 1, 2010, an Appendix R non-compliance issue was identified associated with a potential loss of safety related 2400 VAC Bus 1C and 1D due to specifically sequenced fires affecting control cables and power cables routed in common fire areas but different raceways. An extent of condition corrective action identified that three 2400 VAC breakers 152-103 (Bus 1 C Service Water Pump P-7B), 152-205 (Bus 1D Service Water Pump P-7C), and 152-208 (Bus 1D Component Cooling Water Pump P-52B) all have control cables routed in the same fire areas as the associated power cables. There are three fire areas, the 1C switchgear room 116A (Fire Area 4), screen house room 136 (Fire Area 9), and component cooling water room 123 (Fire Area16) that contain both control cables and power cables for the associated breakers. The concern is that a fire could first damage the control cable(s), resulting in the opening of the control power fuses which would disable the control power and render the breaker protection circuitry nonfunctional. The 2400 VAC breakers fail-as-is on loss of control power. As such if the breakers were closed they would remain closed until manually opened. If the same fire then damaged the power cable(s) on the same breaker, the loss of the breaker protection circuitry would prevent the individual breaker from automatically opening and the clearing of the cable fault would propagate upstream to the next coordinated breaker which would result in the associated 2400 VAC bus feeder breaker opening and a loss of the entire bus. Breaker 152-103 has the issue of both control and power cables present in the screen house room 136 (Fire Area 9) which has potential to result in the loss of the 1C bus for a fire in the screen house. The Appendix R safe shutdown analysis credits both 1C and 1D busses remaining available for a fire in Fire Area 9 and credits 1D bus remaining available for a fire in Fire Area 4. Breaker 152-205 has the issue of both control and power cables present in the screen house room 136 (Fire Area 9) and in the 1C Switchgear room 116A (Fire Area 4) which has potential to result in the loss of 1C and 1D busses for a fire in the screen house or the 1C switchgear room. The Appendix R safe shutdown analysis credits both 1C and 1D busses remaining available for a fire in Fire Area 9 and credits 1D bus remaining available for a fire in Fire Area 4. Breaker 152-208 has the issue of both control and power cables present in the component cooling water room 123 (Fire Area 16) which has the potential to result in the loss of 1D bus for a fire in the component cooling water room. The Appendix R safe shutdown analysis credits both 1C and 1D busses remaining available for a fire in Fire Area 16. The 1C and 1D busses are relied on to provide power to safety related equipment. The busses are normally supplied by offsite power. In the event that offsite power is lost, power is supplied to the busses by the emergency diesel generators. Existing hourly fire tours of the 1C switchgear room, the screen house room, and the component cooling water room are credited as initial compensatory measures. As a compensatory measure, existing hourly fire tours of the 1C switchgear room, the screen house room, and the component cooling water room are credited. This is considered to be an unanalyzed condition that significantly degrades plant safety. The NRC Resident Inspector has been notified. | |
| ENS 46221 | 24 August 2010 02:10:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Loss of Load Trip Signal on All Four Channels of Reactor Protection System | On August 23, 2010, at approximately 2210 hours, with the Palisades unit at 100% power, the turbine protection relays, 305L and 305R, were identified as being inoperable due to a loss of electrical power from their 125 VDC power supply. On an actual turbine low auto stop oil pressure condition, 305L and 305R energize to generate a loss of load trip signal to all four channels of the Reactor Protection System (RPS). 305L provides input to RPS channels A and C. 305R provides input to RPS channels B and D. The RPS generates a reactor trip upon receipt of two (2) of four (4) loss of load trip signals. The loss of load trip is required by Technical Specification (TS) Limited Condition of Operation (LCO) 3.3.1 and was inoperable until 305L and 305R were reenergized on August 24, 2010, at approximately 0939 hours. The on-shift operations crew entered a TS 3.3.1 LCO, however, TS 3.0.3 LCO should have also been entered. TS 3.0.3 LCO was applicable due to all four (4) loss of load input signals to the RPS being inoperable. The eight (8) hour notification was not reported within the required time frame due to a misinterpretation of the event that has since been re-evaluated and determined to meet the eight hour immediate notification requirement. The safety significance of this even was minimal. Although required by TS, the loss of load trip function is not credited in the plant safety analysis. The licensee has informed the NRC Resident Inspector. | |
| ENS 44922 | 21 March 2009 00:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | State of Michigan Notified of Onsite Sewage Spill | At 2030 hours on March 20, 2009 it was determined that sewage from the onsite sewage system had exited the system through a manhole on the 590 foot elevation of the site. This discharge, a clear odorless water, then flowed along the asphalt roadway to a storm drain which ultimately discharges to the beach of Lake Michigan (no fluid reached Lake Michigan waters). The State of Michigan, via the Pollution Emergency Alert System (PEAS), was notified as required by the Site Spill Plan by the Site Environmental Coordinator at 2207 hours. The local government (Van Buren County) was notified at 2220 hours via 911. The licensee informed both state/local agencies and the NRC Resident Inspector. | |
| ENS 44745 | 29 December 2008 11:15:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Ens and Commercial Phone Service | At approximately 0615 on December 29, 2008, it was determined that the Emergency Notification System (ENS) and commercial telecommunications capability were out of service. The condition impacts both the Control Room and the Technical Support Center. The problem has been determined to be an offsite issue with the telecommunications provider. Immediately after discovery of the condition, communications capability via satellite phone was tested and established between the Palisades Control Room and the NRC Operations Center. In addition, if needed, the satellite phone will be used to initiate call-out of the emergency response organization and to contact the regional load dispatcher. This condition did not affect the ENS or commercial capabilities at the offsite Emergency Operations Facility. The telecommunications provider has not provided an estimated repair time. The licensee notified the NRC Resident Inspector.
The telecommunications provider restored full ENS and commercial telecommunications capability to the Palisades Nuclear Plant at approximately 2100 EST. The condition was caused by an offsite fiber-optic cable that had been accidentally severed. The licensee notified the NRC Resident Inspector. Notified the R3DO (Cameron). | |
| ENS 44645 | 7 November 2008 19:41:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Load Calulations for Edg Were Incorrect | During a review it was determined that the Diesel Generator load calculation (EA-ELEC-LDTAB-005) did not account for worst case load from the Containment Air Cooler Fan Motors (V-1A, V-2A, V-3A). Initial review indicates that with the worst case fan motor loading, Diesel Generator 1-2 could be loaded beyond its 2 hour rating following a Loss of Coolant Accident (LOCA) during the time period prior to Recirculation Actuation (RAS). The horsepower loading for the motors used in the Diesel Generator load calculation was based on the original specification for the motors. In 1993 calculation EA-DPAL93-110 was done to determine the impact of partial flooding of the air coolers coils on the cooling fan performance. The 1993 calculation calculated a fan motor load that was higher than what is used in the diesel generator load calculation. However, the diesel generator load calculation was not updated as a result of the 1993 calculation. The higher motor power requirement is 17 kW per fan. There are 3 Containment Air Cooler fans loaded on Diesel Generator 1-2. The extra loading, when combined with the possible additional load from operating the diesel generator at increased frequency, raises the calculated load to 2782 kW, which is above the 2 hour rating of 2750 kW. The overload only applies during a time segment of the diesel generator load profile prior to RAS. This could result in the loss of the diesel generator. There is only 1 Containment Air Cooler Fan Motor (V4A) loaded on Diesel Generator 1-1 and has 121 kW margin available. Therefore, there is no concern about overloading Diesel Generator 1-1. Placed hand switches (42-299CS & 42-277CS) for Turbine Generator Emergency Air Side Seal Oil Backup Pump (P-23) and Turbine Turning Gear Oil Pump (P-26) in Pull-To-Lock position, which prevents the pumps from automatically starting. This reduces the potential load on 1-2 D/G by 71 kW. This restores the load on the diesel to within the margin. Licensee has notified the NRC Resident Inspector
Entergy Nuclear Operations Inc. (ENO) is retracting Event Notification EN #44645 which reported a loss of a safety function due to the 1-2 emergency diesel generator (EDG) load calculation not accounting for the worst-case load from the containment air cooler fan motors. The initial review indicated that, with the worst-case fan motor loading, the 1-2 EDG could have been loaded beyond the 2750 kW two-hour peak loading limits following a loss-of-coolant accident during the time period prior to a recirculation actuation signal. This condition may have caused 1-2 EDG to become inoperable, and could have prevented fulfillment of a safety function. In a subsequent evaluation of loading capability, completed on December 18, 2008, ENO determined the 1-2 EDG was operable. The EDGs are rated for 2750 kW two-hour peak operation. It was determined that the postulated peak load for the 1-2 EDG would have been 2819 kW for a period of approximately thirty-eight minutes. However, based on engineering information obtained from the vendor of the 1-2 EDG, Fairbanks-Morris Engine, and reviewed by ENO, the 1-2 EDG could have been operated up to 2830 kW for fifty minutes before any susceptibility to damage might occur. In addition, under the peak loading condition of 2819 kW, the speed of the 1-2 EDG would have remained above the Technical Specification limit of 59.5 Hz. The subsequent review confirmed that the safety function would have been fulfilled. Therefore, ENO is retracting this event notification. The licensee has notified the NRC Resident Inspector. R3DO (Cameron) notified. | |
| ENS 44533 | 2 October 2008 13:15:00 | 10 CFR 26.719, FFD Reporting requirements | Positive Fitness for Duty Test | A non-licensed contract employee supervisor had a confirmed positive for alcohol during a follow up fitness-for-duty test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | |
| ENS 44385 | 6 August 2008 00:29:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Tech Spec Required Shutdown Due to Excessive Unidentified Leakage | On August 5, 2008, a planned shutdown was in progress to repair a control rod drive mechanism due to excessive leakage (within Tech Spec identified leakage limits). During the shutdown, the site experienced excessive unidentified leakage of approximately four gallons per minute for short periods following charging pump starts, due to leakage past the letdown relief valve. (this first occurred at 1629) As a result of the plant shutdown, and not being in a steady state condition, PCS leakage is unable to be definitively determined. Limiting Condition of Operation 3.4.13 was entered at 1629 hrs. The plant was in Mode 3 at 1955 hrs. The Tech Spec requirement is to restore leakage to within limits within 4 hours. If not within limits within 4 hours, be in Mode 3 within 6 hours (August 6, 2008 at 0229 hours) and be in Mode 5 within 36 hours (August 7, 2008 at 0829 hrs). This event is being reported as a 4-hour non-emergency report in accordance with 10 CFR 50.72(b)(2)(i), Tech Spec required shutdown. The licensee will notify the NRC Resident Inspector. | |
| ENS 44237 | 23 May 2008 16:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Possible Generator Lockout | The reactor automatically tripped at 1249 EDT on 05/23/08. The cause of the reactor trip signal is not positively known, but it is believed to be related to trip of the 346 negative sequence relay which caused a trip of the 386C generator independent trip (coastdown) lockout relay. All plant systems performed as designed with minor deficiencies noted. EOP-1, 'Standard Post Trip Actions', and EOP-2, 'Reactor Trip Recovery', performed. All control rods fully inserted. All atmospheric steam dump valves lifted and reseated. CV-0780 (B Steam Generator ASDV) was identified as leaking and was manually isolated. There is no known primary to secondary leakage. Auxiliary feedwater system actuated automatically as designed in response to low steam generator water levels. AFW actuation consisted on AFW pump P-8A auto start and AFW valve operation. Steam generator water levels were subsequently restored to normal and are being maintained for primary coolant system heat removal via main steam isolation valves to main condenser (via the turbine bypass valves). Plant is at Normal Operating Pressure (2060 psia) and Temperature (533 degrees F). Grid is stable with safety bus station power being provided from Safeguards transformer. The licensee notified the NRC Resident Inspector. |