Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 55113 | 24 February 2021 18:22:00 | On February 23, 2021, it was discovered that two licensed operators had manipulated the plant between April 30, 2020 and July 20, 2020, without being fully qualified; having exceeded the due date for their annual operating test. This constitutes a deviation from the facility's Technical Specification 6.1.3.1 - 'Minimum staffing'. During that period the reactor was operated normally, there were no unusual occurrences and at no time did the loss of qualifications cause the existence or development of an unsafe condition with regard to reactor operations. Both operators performed and passed their operating test on July 20, 2020. The cause of this event is under investigation. The licensee has notified their NRC Project Manager. |
| ENS 53575 | 31 August 2018 16:04:00 | The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. This event is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) as a Loss of Emergency Preparedness Capabilities at Palo Verde Nuclear Generating Station (PVNGS) Unit 2. On August 31, 2018, at approximately 0544 Mountain Standard Time (MST), the Unit 2 control room experienced an unplanned loss of Steam Generator #1 steam line monitor (RU-139), Channels A and B. This main steam line monitor is used in the PVNGS Emergency Plan to perform dose assessment in the event of a steam generator tube rupture. The NRC Resident Inspectors have been notified. |
| ENS 53424 | 23 May 2018 17:37:00 | The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. On May 23, 2018, at approximately 1128 Mountain Standard Time (MST), the Palo Verde Nuclear Generating Station (PVNGS) Unit 2 control room received reactor protection system alarms for low departure from nucleate boiling ratio and an automatic reactor trip occurred. Prior to the reactor trip, Unit 2 was operating normally at 100 percent power. Plant operators entered the reactor trip procedures and diagnosed an uncomplicated reactor trip. All CEAs (control element assemblies) fully inserted into the core. No emergency classification was required per the PVNGS Emergency Plan. The Unit 2 safety-related electrical buses remained energized from normal offsite power during the event. There was no impact to the required circuits between the offsite transmission network and the onsite Class 1E Electrical Power Distribution System; the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 2 is currently stable in Mode 3 with the reactor coolant system at normal operating temperature and pressure. The cause of the reactor trip is under investigation. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The NRC Resident Inspector has been informed of the Unit 2 reactor trip. Decay is being removed via steam dumps to condenser. Units 1 and 3 at Palo Verde were unaffected by the transient and continue to operate at 100 percent power. |
| ENS 52670 | 7 April 2017 14:49:00 | On April 5, 2017, Arizona Public Service Company (APS) completed an evaluation of a deviation, and concluded the condition represented a defect under 10 CFR 21. APS previously submitted an interim report (ADAMS Accession Number ML 16344A118) for this condition pursuant to 10 CFR 21.21(a)(2). A GE-Hitachi Type AKR-2BE-50, 2000 Amp circuit breaker (used to connect Class 1E batteries to the related Class 1E 125 VDC busses) exhibited arcing and smoking during current injection testing performed to test the overcurrent trip setpoint prior to installation. Arcing occurred at one of two hex bolts anchoring the protective trip device to the line side bus. The electrical arcing resulted from inadequate tightening of both hex bolts which caused a loose electrical connection on the bus within the breaker. APS concluded this condition could result in the breaker failing to perform its safety function and thus could create a substantial safety hazard. The breaker had been refurbished by GE-Hitachi and was received by APS and tested on October 13, 2016. Following the test failure, the damaged bolt was replaced, both bolts were tightened, and the breaker was retested and installed. Pre-installation inspection and testing that includes current injection testing, recommended in GEH document GEK-64459, should, and did, detect faults such as the condition identified in this notification. GE-Hitachi entered this failure into their corrective action program. Vendor: GE HITACHI NUCLEAR ENERGY, 3901 CASTLE HAYNE RD., WILMINGTON, NC 28402-2819 Device: Breaker Model AKR-2BE-50, 2000 Amp, Serial No. N8682600001 The NRC Resident Inspector has been informed. |
| ENS 51910 | 5 May 2016 15:39:00 | The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. At 0500 MST on May 5, 2016, engineering personnel determined that leakage from the bonnet seal weld of the Train B High Pressure Safety Injection (HPSI) system loop 2A injection valve, SIBUV0616, constituted reactor coolant system pressure boundary leakage. This is being reported as a degradation of a principal safety barrier pursuant to 10 CFR 50.72(b)(3)(ii)(A). The leak was identified during a planned activity in which Operations was filling the refueling pool using HPSI pump B. Leakage was stopped when a plant operator closed SIBUV0616. PVNGS Unit 1 was shut down for its 19th refueling outage (1R19) on April 9, 2016, at 0000 MST and is in Mode 6. The NRC Resident Inspectors have been informed of this condition.
Subsequent engineering evaluation of the leak condition concluded the leakage from the bonnet seal weld of SIBUV0616 was not reportable reactor coolant system pressure boundary leakage. This conclusion was based on further reviews of the PVNGS licensing bases, ASME Code requirements, and design features of the valve with vendor assistance, which determined that the body-to-bonnet threads provide the structural support for the mechanical joint and the seal weld is not required for structural integrity of the component. Based on the above information, PVNGS has determined the leakage did not represent a degraded condition of a principal safety barrier as defined by 10 CFR 50.72(b)(3)(ii)(A) and Event Notification No. 51910 is hereby withdrawn. The NRC Resident Inspectors have been informed. Notified R4DO (Werner) |
| ENS 50021 | 10 April 2014 13:52:00 | The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. This notification describes a previously undeclared event that is now reported pursuant to the requirements of 10 CFR 50.72(a)(3). No emergency situation exists at this time. On November 6, 2013, while operating at 100% reactor power under steady state conditions, a U-1 control room alarm was received indicating low fluid pressure in the Train 'A' hydraulic accumulator of Main Steam system isolation valve MSIV-170. A plant operator was dispatched according to the alarm response procedure to perform a pre-charge check of the accumulator. A pre-charge check returns hydraulic fluid from the accumulator to a vented tank that serves as a fluid reservoir. During this operation the fluid reservoir tank and supporting piping were damaged by the rapid expansion of nitrogen gas that had entered the hydraulic system by way of a leak across an accumulator piston O-ring. As a result of the damage and loss of hydraulic fluid, MSIV-170 and its associated hydraulic accumulators were declared inoperable at 1307 Mountain Standard Time. It was determined at that time that the failure of the fluid reservoir due to rapid expansion did not constitute an explosion and a declaration of an Unusual Event was not made according to Emergency Action Level (EAL) HU2, 'Fire within the Protected Area not extinguished within 15 minutes of detection or explosion within the Protected Area.' Subsequent review has determined that the reservoir failure should have been characterized as an explosion within the Protected Area and that a declaration of an Unusual Event should have been made per EAL HU2. The NRC Resident Inspector has been notified. The licensee has notified the State of Arizona and Maricopa County. |