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 Entered dateSiteRegionReactor typeEvent description
ENS 538199 January 2019 13:23:00PalisadesNRC Region 3At 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 538133 January 2019 23:57:00PalisadesNRC Region 3At 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 5374921 November 2018 17:27:00PalisadesNRC Region 3On 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 5373411 November 2018 21:59:00PalisadesNRC Region 3

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 5373310 November 2018 17:48:00PalisadesNRC Region 3On 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 5276419 May 2017 08:48:00PalisadesNRC Region 3CEPursuant 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 527222 May 2017 15:26:00PalisadesNRC Region 3CE

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.

  • * * RETRACTION ON 5/9/17 AT 1303 EDT FROM BARBARA DOTSON TO BETHANY CECERE * * *

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 5264729 March 2017 18:10:00PalisadesNRC Region 3CEDuring 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 5192212 May 2016 13:35:00PalisadesNRC Region 3CEA 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 5182024 March 2016 08:58:00PalisadesNRC Region 3CEAt 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 5139716 September 2015 03:59:00PalisadesNRC Region 3CEAt 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 5122413 July 2015 13:33:00PalisadesNRC Region 3CEOn 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 510332 May 2015 12:58:00PalisadesNRC Region 3CE

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.

  • * * UPDATE FROM JC RANEY TO DANIEL MILLS AT 1601 EDT ON 5/2/15 * * *

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 5090719 March 2015 16:44:00PalisadesNRC Region 3CEOn 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 501008 May 2014 21:57:00PalisadesNRC Region 3CEAt 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 4996026 March 2014 10:55:00PalisadesNRC Region 3CEThis 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 4995926 March 2014 10:55:00PalisadesNRC Region 3CEThis 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 4992417 March 2014 15:12:00PalisadesNRC Region 3CEAt 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 497965 February 2014 16:17:00PalisadesNRC Region 3CEOn 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 4977329 January 2014 11:06:00PalisadesNRC Region 3CEOn 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 495167 November 2013 16:32:00PalisadesNRC Region 3CEA 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 4929822 August 2013 15:05:00PalisadesNRC Region 3CEA 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 4927613 August 2013 16:11:00PalisadesNRC Region 3CEAt 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 490025 May 2013 01:45:00PalisadesNRC Region 3CEAt 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 4875815 February 2013 15:05:00PalisadesNRC Region 3CEAt 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 484784 November 2012 14:52:00PalisadesNRC Region 3CEAt 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 4818212 August 2012 11:05:00PalisadesNRC Region 3CEFollowing 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 4801812 June 2012 17:27:00PalisadesNRC Region 3CE

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.

  • * * RETRACTION ON 8/7/12 AT 1219 EDT FROM TERRY DAVIS TO DONG PARK * * *

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 478158 April 2012 05:01:00PalisadesNRC Region 3CENotified 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 4752314 December 2011 16:40:00PalisadesNRC Region 3CEThe 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 473225 October 2011 18:00:00PalisadesNRC Region 3CEAn 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 4729025 September 2011 18:21:00PalisadesNRC Region 3CEAt 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 4727116 September 2011 15:04:00PalisadesNRC Region 3CE

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 4718723 August 2011 15:07:00PalisadesNRC Region 3CE

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.

  • * * UPDATE FROM TODD MULFORD TO JOHN KNOKE AT 1842 EDT ON 8/23/11 * * *

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 4662719 February 2011 22:24:00PalisadesNRC Region 3CEAt 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 4656422 January 2011 18:24:00PalisadesNRC Region 3CEThe 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 465248 January 2011 13:21:00PalisadesNRC Region 3CE

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.

  • * * UPDATE FROM CARYLIN MCCOY TO JOHN KNOKE AT 0102 EST ON 01/09/11 * * *

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 4650522 December 2010 22:12:00PalisadesNRC Region 3CEA 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 463021 October 2010 15:31:00PalisadesNRC Region 3CEAt 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 462211 September 2010 15:34:00PalisadesNRC Region 3CE

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 4492220 March 2009 22:56:00PalisadesNRC Region 3CEAt 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 4474529 December 2008 11:19:00PalisadesNRC Region 3CE

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.

  • * * UPDATE AT 2104 EST ON 12/29/08 FROM MULFORD TO HUFFMAN * * *

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 446457 November 2008 22:00:00PalisadesNRC Region 3CE

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

  • * * RETRACTION AT 1416 EST ON 12/30/08 FROM DAVIS TO HUFFMAN * * *

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 445332 October 2008 12:03:00PalisadesNRC Region 3CEA 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 443855 August 2008 22:27:00PalisadesNRC Region 3CEOn 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 4423723 May 2008 16:06:00PalisadesNRC Region 3CEThe 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.
ENS 4414818 April 2008 06:23:00PalisadesNRC Region 3CE

The earthquake in southeastern Illinois was felt on site. No damage has been reported. All secondary system walkdowns have been completed. Primary system walkdowns are in progress. The licensee notified the NRC Resident Inspector.

* * * UPDATE FROM P. PITCHER TO P. SNYDER AT 1144 on 4/18/07 * * *

Palisades continues in the Unusual Event after detecting an aftershock from the previous earthquake. Palisades is continuing checks per their procedures for the circumstances at the site.

  • * * UPDATE FROM TODD PLAUGHER TO JASON KOZAL AT 1438 on 4/18/07 * * *

The licensee has exited the Unusual Event at 1431 on 4/18/08. The licensee notified the NRC Resident Inspector. Notified R3DO (Kozak), NRR EO (Ross-Lee), IRD (McDermott), DHS (Gomez), FEMA (Burckhardt)

ENS 4405913 March 2008 12:38:00PalisadesNRC Region 3CEAbstract: Palisades has experienced failures of two snubber valves on an emergency diesel generator since November 2005. These valves were made from AISI E 52100 material. Entergy has determined through destructive metallurgical analysis that the material was incorrectly heat-treated. Not all of the install snubber valves were constructed of this material. The defect in the material of the snubber valves may allow the snubber valves to crack under operation, resulting in fuel oil leakage and loss of fuel oil supply to the affected cylinder on the engine. Component information: Manufacturer: Fairbanks Morse Model: ALCO Model 251F ALCO Part Number: 2402466 Description: Palisades has experienced failures of two snubber valves on the 1-2 emergency diesel generator since November 2005 due to incorrectly heat treated material of the snubber valves. Both snubber valves failed soon after being initially installed in the engine or after being re-installed in the engine after maintenance. The first snubber valve failed after operating for approximately 10 hours of operation. The second snubber valve was installed in the engine and had approximately 135 hours of run time on it before it was removed, inspected and re-installed in the engine. After reinstallation into the engine, the snubber valve failed after approximately 2 hours of operation. Review of the purchasing history of the snubber valves could not identify when these failed snubber valves were originally purchased. On February 27, 2008, Entergy determined, through destructive metallurgical analysis, that the material that was incorrectly heat-treated was AISI E52100 material. Not all snubber valves were constructed of this material. Subsequently, the receipt inspection practices were revised to include testing of the snubber valves for the suspect material. This is accomplished by using a Niton XLT 800 chemical analyzer to identify (non-destructively) the chemical content of the snubber valves. Snubber valves with manganese contents <0.5% are rejected as they are most likely to be constructed of E52100 material. In January 2008, five snubber valves were rejected due to not meeting Palisades material requirements. Subsequently, one of the snubber valves was destructively tested and confirmed to have been incorrectly heat treated. These snubber valves were identified to have been purchased in 2004. A Part 21 evaluation was performed and it was concluded that this is a substantial safety hazard because the defect may allow the snubber valves to crack under operation, resulting in fuel oil leakage and loss of fuel oil supply to the affected cylinder on the engine. The loss of fuel oil supply to a cylinder may affect the ability of the emergency diesel generator to meet its design basis load rating; the external leakage of fuel may result in the potential for a fire to develop and render the diesel generator inoperable; and the leakage of fuel may affect the ability of the fuel oil system to supply adequate fuel oil to the diesel generator throughout its mission time. Causes: Snubber valves constructed from AISI E52100 material may be incorrectly heat-treated. The cause of this is not known. Corrective Actions: Palisades experienced two failures, which were reported in Licensee Event Reports 2005-007 and 2007-006. Entergy identified the improper heat treatment following the second failure. Entergy has subsequently replaced all snubber valves on the 1-1 emergency diesel generator. Entergy replaced the defective snubber valve and the remaining snubber valves on the 1-2 emergency diesel generator that had been replaced at the same time the defective snubber was installed. Since the replacement snubber valves on 1-2 emergency diesel generator did not have their material composition verified, eight additional snubber valves were sub replaced after verifying E52100 was not present. Entergy plans to replace the remaining 10 snubber valves on 1-2 emergency diesel generator. Based on the amount of in-service time of these snubber valves and verification of 26 acceptable snubber valves, there is reasonable expectation of operability. The licensee notified the NRC Resident Inspector. Palisades has been in contact with Fairbanks Morse concerning this issue.
ENS 4390013 January 2008 09:11:00PalisadesNRC Region 3CEPalisades Operations staff manually tripped the reactor in accordance with their station's Off-Normal Operating Procedure, after an automatic trip of the "B" Main Feedwater Pump (MFP). The cause of the MFP auto-trip is unknown and being investigated. During the trip, all systems performed as designed. During the transient resulting from the trip, the atmospheric steam dump valves did open momentarily and properly reseated. There is no known primary-to-secondary leakage therefore no release was known to have occurred. All control rods inserted fully as a result of the trip. Auxiliary feedwater pump (AFW) 8A started, as expected, in response to low steam generator water levels experienced during the transient and supplied both steam generators. Decay heat is being removed via the main steam isolation valves to the condenser with the 8A AFW pump supplying feedwater to the steam generators. The plant is at normal temperatures and pressures for this mode. The electrical grid is stable with the station start-up transformer supplying station loads. The licensee has notified the NRC Resident Inspector.
ENS 4383210 December 2007 22:21:00PalisadesNRC Region 3CEFive new ground water monitoring wells were recently installed at Palisades Nuclear Plant in support of the Nuclear Energy institute (NEI) ground water initiative. The initial sampling of one of these wells displayed a level of tritium that triggered the communication protocol of the NEI initiative on ground water protection. On December 10, 2007, at 1830 hours, Entergy confirmed that the tritium concentration for this well was 22,000 picoCuries per liter (pCi/l). The threshold for initiating the communication protocol is 20,000 pCi/l (Offsite Dose Calculation Manual limit for drinking water). This well is located inside the owner controlled area and inside the protected area. This well is not a drinking water source. Entergy is continuing to investigate the source of tritium identified in this well. Samples from the remaining four wells are below minimum detectable activity levels. There is no indication that tritium has migrated off the Palisades site. The licensee plans to notify the State of Michigan, Van Buren County Office of Domestic Preparedness, City of South Haven, Covert Township, and the South Haven Charter Township. The licensee has notified the NRC Resident Inspector.