|Report date||Site||Event description|
|05000255/LER-2017-002||17 July 2017||Palisades|
On May 19, 2017, at 0206 hours, an unexpected Reactor Protection System (RPS) actuation occurred during pre-startup testing. The reactor was shutdown at the time, with all control rods inserted. The portion of the test that was in progress is designed to actuate the RPS from a loss of load input signal. To facilitate this part of the test with the reactor in a shutdown mode, one of two conditional steps in the procedure is to be taken. The generator motor operated disconnect 389 (MOD-389) is required to be in the open position, or protective trip circuity for the generator is required to be bypassed.
Due to a conditional step of the test procedure being misinterpreted by a Nuclear Control Operator (NCO), MOD-389 was left in the closed position and the generator protective trip circuity was not bypassed. This resulted in the RPS actuation occurring prior to the preplanned sequence. The RPS responded as designed. All components operated as expected for the plant conditions.
The cause of the unexpected RPS actuation was human performance errors during procedure performance, e.g., lack of self-validation/verification, misinterpretation of information, and lack of peer check verification.
Corrective actions from the event include the removal of the NCO's licensed operator qualifications until remediated and initiation of a standing order requiring peer check verification for all procedure conditional steps until the applicable administrative procedure is revised. Additionally, a case study of the event will be included in a 2017 operations high intensity training session.
|05000255/LER-2017-001||24 May 2017||Palisades|
On March 29, 2017, during an evaluation of protection of Technical Specification (TS) equipment from the damaging effects of tornados, nonconforming conditions were identified in the plant design. Specifically, TS equipment did not meet current design basis for protection against potential tornado missile impact. Identified components/systems were declared inoperable and NRC Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado Generated Missile Protection Noncompliance," was implemented. Initial compensatory measures were implemented, per the guidance of NRC Interim Staff Guidance DSS-ISG-2016-01 Appendix A, within the time allowed by the applicable Limiting Conditions for Operation (LCOs) and the associated systems were then declared operable but nonconforming.
The six systems, containing TS required equipment, did not meet current design basis for protection against potential tornado missile impact. Credible tornado missile impacts could affect the following systems; Service Water, Fuel Oil, Emergency Diesel Generators, Auxiliary Feedwater, Component Cooling Water and Control Room Ventilation Filtration.
Comprehensive compensatory measures will be implemented in approximately 60 days of discovery, per the guidance of NRC Interim Staff Guidance DSS-ISG-2016-01 Appendix A.
Due to the historical nature of the issue, a specific cause for the identified vulnerabilities was not determined.
|05000255/LER-2015-001||10 November 2015||Palisades|
On September 16, 2015, at approximately 0117 hours, an anomaly within the digital electro-hydraulic (DEH) turbine control system initiated a turbine trip. As designed, the turbine trip actuated the reactor protection system to automatically trip the reactor due to a loss of load and the auxiliary feedwater system started automatically to recover steam generator levels.
The direct cause of the event is the turbine tripped due to actuation of the "DEH controller loss of power" turbine trip logic. Troubleshooting and analysis determined there was a failure of a power supply module on a circuit board in the DEH turbine control system. Subsequent to the power failure on the circuit board, a second failure, either a loss of power to the overspeed protection control (OPC) distributed processing units (DPUs) or a loss of communications between the primary and backup OPC DPUs, occurred resulting in an actuation of the "DEH controller loss of power" turbine trip logic.
The root cause of the event is that the design of the DEH system contains unnecessary trip logic associated with turbine overspeed monitoring. Corrective actions include a modification to remove the DEH system OPC loss of power and loss of communications trip logic.
|05000255/LER-2013-003||11 October 2013||Palisades|
At 1102 on August 13, 2013, both control room ventilation filtration system trains were declared inoperable in accordance with Technical Specification (TS) 3.7.10, Condition B, due to the inability to fully close control room envelope (CRE) boundary door-15. At 1111 on August 13, 2013, door-15 was closed and TS 3.7.10, Condition B, was exited. TS 3.7.10 allows CRE boundary doors to be opened intermittently, under administrative control for preplanned activities, provided the doors can be rapidly restored to the design condition.
During preplanned maintenance activities, workers attempting to exit the CRE area were unable to open door-15 via normal operation of the door's hand wheel. Recent frequent operation of door-15 may have caused deformation of a cotter pin within the door's normal operating mechanism. Deformation of the cotter pin could cause the normal operation of door-15 to function intermittently. To allow exiting, workers opened door-15 using the emergency egress latch that activated an alarm condition on the door. During exiting, inadvertent operation of the door's hand wheel in the closed direction caused the door's latching pins to extend out causing interference between the door and the door frame preventing door-15 from fully closing.
Due to the door being in an alarmed condition, the door's latching pins were unable to be immediately retracted. After approximately nine minutes, the door's latching pins were retracted by use of the emergency egress latch and the door was restored to the design condition, i.e., closed.
The cotter pin was replaced. Future potential corrective actions include increased preventative maintenance frequency to replace the cotter pin and restricting the use of the emergency egress latch.
|05000255/LER-2013-002||25 June 2013||Palisades|
At 0027 on May 5, 2013, the safety injection/refueling water (SIRW) tank was declared inoperable in accordance with the operational decision-making issue (ODMI) process. Water leakage from the tank had exceeded the pre-established limit of the ODMI process that directed the tank be declared inoperable.
Leakage from the tank was quantified at approximately ninety gallons per day. Technical Specification (TS) 3.5.4.B requires restoration of an inoperable SIRW tank within one hour. If the tank is not returned to an operable status within one hour, TS 18.104.22.168 requires the plant be placed in Mode 3 within six hours and in Mode 5 within the subsequent thirty-six hours.
Due to the inability to repair the leak within the required one-hour time frame, a plant shutdown was initiated at approximately 0100 hours on May 5, 2013. The plant entered Mode 3 at 0457 hours on May 5, 2013. At 2358 hours on May 5, 2013, the plant entered Mode 5 to execute repairs.
Testing identified an approximate 3/16-inch through-wall crack in a nozzle reinforcing collar to floor plate weld of the tank. Follow-up analysis determined there was significant lack of fusion in the weld that resulted in the failure of the weld and subsequent water leakage. The welder that fabricated the weld did not ensure adequate fusion at the weld root.
The entire SIRW tank floor was replaced with the exception of an annulus ring around the perimeter.
Several initiatives were implemented to preclude potential weld issues during the fabrication of the new tank floor, including welder proficiency training on revised welding techniques and utilization of several types of weld testing methods.
|05000255/LER-2002-002||21 January 2003||Palisades|
On December 1, 2002, at approximately 2154 hours, with the plant operating at 100% power, an automatic reactor trip occurred on main generator loss of load. The loss of load occurred when a transmission tower's static line hanger failed, allowing one of two static lines to contact a 345 KV transmission line, tripping the main generator. The static line also contacted the rear bus in the switchyard that supplies the plant non-1E 4160 volt startup transformers. The rear bus tripped on a fault-to-ground causing a loss of non-1E 4160 volt AC buses. Consequently, both main feedwater pumps tripped, and the auxiliary feedwater system started automatically on low steam generator level, as expected.
The plant was maintained at or near normal operating pressure and temperature subsequent to the trip, on natural circulation, since startup power for primary coolant pumps was also lost. The plant was returned to service on December 5, 2002.
This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an automatic reactor trip and automatic actuation of the auxiliary feedwater system.
|05000255/LER-2008-004||5 August 45 JL||Palisades|
On July 15, 2008, with the plant in Mode 1 at 100% power, Entergy Nuclear Operations, Inc. personnel discovered that the Palisades Nuclear Plant region 1 spent fuel pool (SFP) storage racks contain less neutron absorber material than assumed in the SEP criticality analysis of record. This neutron absorber material is relied on to maintain the region 1 SFP storage racks within the Technical Specification (TS) 22.214.171.124.b criticality requirements. � The TS reflects credit for the neutron absorber material in maintaining SFP criticality within limits. At the time of discovery, SFP boron concentration was 2732 ppm.
� TS 126.96.36.199.b requires that Keff for region 1 fuel racks be less than or equal to 0.95 if fully flooded with unborated water.
With soluble boron required to maintain Keff less than or equal to 0.95 in the region 1 fuel racks, assuming nominal enrichment, PNP no longer complies with TS 188.8.131.52.b.
The degraded neutron absorber material did not involve an immediate safety concern at the time of discovery because the SFP boron concentration was 2732 ppm, and a SEP criticality operability assessment concluded that a soluble boron concentration of 150 ppm is required to maintain a Keff less than or equal to 0.95 in the region 1 racks. � In addition, plant procedures required that SFP boron concentration be maintained at a minimum of 2550 ppm in Modes 1 through 4. In Modes 5 and 6, 1800 ppm was required. A past operability evaluation confirmed that SFP boron concentration had been maintained greater than 2550 ppm in recent years. Compensatory measures are in place. This condition does not represent a safety system functional failure. This is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS.