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| Start date | Title | Description | Topic | |
|---|---|---|---|---|
| ENS 56501 | 2 May 2023 19:00:00 | 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 52722 | 2 May 2017 13:28:00 | 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). | Time of Discovery |
| ENS 48018 | 12 June 2012 18:56:00 | 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 44645 | 7 November 2008 19:41:00 | 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 43186 | 25 February 2007 18:00:00 | Degraded Electrical Cables to Ccw System | On February 17, 2007, it was discovered that cables in a cable tray associated with the Component Cooling Water (CCW) system and Service Water system had sustained external damage by the heat effects from an un-insulated pipe that was in close proximity to the cable tray. On February 25, 2007, at 1100 hrs, as a result of continued evaluation of the cables in the cable tray and the discovery of a cable with unacceptable cable damage, it was determined that all cables located within the cable tray at the effected zone (that were not already isolated or replaced) were inoperable due to the loss of qualification life and the potential for cable-to-cable interaction from degradation of the cable insulation. This condition could potentially result in spurious equipment operation, causing components to be positioned in other than their desired safety position. Subsequently, on February 25, 2007, at 1300 hrs, it was recognized that a specific combination of two postulated cable faults could affect the CCW system resulting in a condition where there would be less than 100% of the required CCW post accident cooling capability, per Technical Specification (TS) 3.7.7.C. As a result, TS 3.0.3 was entered. At 1322 hrs, TS 3.0.3 was exited, following isolation of one of the two postulated cable faults, which restored the capability for 100% of the required CCW post accident cooling. The condition of potentially having less than 100% of the required CCW post accident cooling is reportable per 10 CFR 50.72(b)(3)(v)(B) as a condition that could have prevented the fulfillment of the safety function of structures or systems (CCW system) that are needed to remove residual heat. The licensee notified the NRC Resident Inspector.
EN #43186 reported on February 21, 2007, that cables associated with the component cooling water system had sustained external damage by the heat effects from an uninsulated pipe in close proximity to the cable tray. This was reported as a condition that could have prevented the fulfillment of the safety function of the component cooling water system, which is needed to remove residual heat. Subsequent evaluation of past operability concluded that these cables would not have failed as postulated, and are considered to have been operable. Therefore, the condition did not prevent fulfillment of a safety function and EN #43186 is being retracted. The licensee notified the NRC Resident Inspector. Notified R3DO (S. Burgess). | Past operability |
| ENS 42959 | 2 November 2006 19:36:00 | High Pressure Safety Injection (Hpsi) Pumps Alignment Blocks Improperly Installed | At 1436 hours on November 2, 2006, with the plant in Mode 3, it was determined that less than 100% of the required Emergency Core Cooling System (ECCS) flow was available per Technical Specification (TS) 3.5.2.D. Therefore, TS LCO 3.0.3 was entered. Each High Pressure Safety Injection (HPSI) pump (one in each ECCS train) is designed with alignment blocks in its mounting, which ensures pump and motor alignment for the thermal expansion experienced by the pump upon initiation of sump recirculation flow. These alignment blocks ( 2 per pump) were discovered to be improperly installed, or missing altogether. In this condition, the HPSI pump could potentially be rendered inoperable upon initiation of sump recirculation. This condition is reportable in accordance with 10CFR 50.72(b)(3)(ii)(B) and (b)(3)(v)(D) as an unanalyzed condition, and a condition that could have prevented the fulfillment of the safety function of the HPSI pumps to mitigate the consequences of an accident, respectively. The NRC Resident Inspector was notified of this event by the licensee.
EN # 42959 reported on November 2, 2006 that both Emergency Core Cooling System Trains were inoperable. The reason for that determination involved the observation that the alignment blocks (keys) associated with the mounting of the high pressure safety injection (HPSI) pumps were either incorrectly installed or were missing. The alignment keys were believed to be necessary to ensure appropriate HPSI pump and motor alignment for the thermal expansion experienced upon initiation of sump recirculation flow. The condition was reported as an unanalyzed condition and a condition that could have prevented the fulfillment of the safety function of the HPSI pumps to mitigate the consequences of an accident. Subsequently, further evaluation of the HPSI pump mounting configuration determined that the alignment keys are not required for pump operability. Therefore, there was no unanalyzed condition and no condition that could have prevented the fulfillment of the safety function of the HPSI pumps to mitigate the consequences of an accident. The licensee notified the NRC Resident Inspector. R3DO(Cameron) notified. | Unanalyzed Condition |
| ENS 41128 | 16 October 2004 21:30:00 | Reactor Vessel Head Nozzle Cracking | On 10/16/04, with the Plant in Mode 6, Nuclear Management Company identified through wall cracks in the inconel buttering adjacent to the J-weld on reactor head penetrations 29 and 30. Ultrasonic examinations of the reactor vessel head were being performed in accordance with the First Revised Order, EA-03-009. The ultrasonic examinations identified leak path detection indications. Therefore, in accordance with the Order, a bare metal visual inspection of the exterior of the reactor head was performed. No evidence of leakage was discovered. A dye-penetrant examination was then performed. The dye-penetrant exam showed minor surface indications that required further examination. Following minor excavation of the weld surface, the through wall cracks were identified. This condition was determined to be reportable at 1730 EST for penetration 30, and at 1915 EST for penetration 29. This notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A). Reactor head penetrations 29 and 30 will be repaired prior to placing the reactor vessel head back in service. The licensee notified the NRC Resident Inspector.
On October 17, 2004, Nuclear Management Company notified the NRC, via non-emergency event report 41128 that leak path detection indications had been identified in two reactor pressure vessel head control rod drive mechanism nozzle penetrations at Palisades. At that time, the probable cause of the identified indications was believed to be primary water stress corrosion cracking (PWSCC). As a result, event report 41128 was made in accordance with 10 CFR 50.72(b)(3)(ii)(A) as a condition that resulted in a principle safety barrier being seriously degraded. Based on a recently completed engineering evaluation that compared the indications found in 2004 to the indications identified in subsequent examinations, it was determined the indications originally discovered in 2004 were embedded welding indications, caused by the original welding process. Since the indications identified in 2004 were not caused by PWSCC, serious degradation of a principle safety barrier did not exist. Consequently, the reporting criterion of 10 CFR 50.72(b)(3)(ii)(A) is not applicable. Therefore, event report 41128 is being retracted. In addition, Palisades Licensee Event Report (LER), 2004-002, 'Leak Path Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations,' is being cancelled. The licensee has notified the NRC Resident Inspector. Notified R3DO (Peterson). |