Semantic search
| Start date | Reporting criterion | Event description | |
|---|---|---|---|
| 05000261/LER-2017-002 | 24 October 2017 14 December 2017 | On October 24, 2017 at 1350 EDT, with the plant in Mode 1 at 100 percent power, H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2), discovered an uncompensated security vulnerability that could have allowed unauthorized or undetected access to a vital area (VA) of the plant. Compensatory measures were promptly established and an immediate investigation was conducted, which did not reveal any evidence of unauthorized entry or tampering. The cause of this event is attributed to the absence of work instructions to contact security to establish compensatory measures prior to beginning work activities high on a VA barrier wall. The vulnerability was the result of authorized work. All workers involved were validated to have proper permissions to access the area. No unauthorized access to the VA occurred during the time period that the vulnerability existed. This event did not have any nuclear or personal safety implications. This report is being submitted in accordance with the requirements of 10 CFR 73.71(d), which refers to Appendix G, "Reportable Safeguards Events," Section 1(c). | |
| 05000261/LER-2017-001 | 3 April 2017 1 June 2017 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | At 2155 hours Eastern Daylight Time on 4/3/2017 with the plant in Mode 3 at zero percent power, H. B. Rob.nson Steam Electric Plant, Unit No. 2 (FIBRSEP2), experienced an actuation of the Auxiliary Feedwater (AFW) System during turbine trip logic surveillance testing. Subsequent investigation determined that the surveillance test was performed without verifying AFW actuation signals were defeated, as required by the test procedure. During performance of the test the AFW system actuated when the only running main feedwater (MFW) pump was tripped as part of the test. Since the AFW defeat switches were not in the defeat position, the AFW system actuated as designed in response to the tripped MFW pump. Main feedwater was restored and the AFW pumps were secured. The direct cause of the AFW system actuation was inadequate procedure adherence during turbine trip surveillance testing. |
| 05000261/LER-2016-005 | 8 October 2016 22 February 2017 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | At 1302 hours Eastern Daylight Time (EDT) on 10 08 2016 with the plant in Mode 1 at 100 percent power, H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2), experienced a grid perturbation. As a result, HBRSEP2 experienced a reactor trip due to low voltage on the 4kV buses. Plant safety systems responded with the emergency buses separating from offsite power due to emergency bus undervoltage. The emergency diesel generators (EDG) started and powered the 480V emergency buses. 'A' service water pump did not start on the blackout sequencer; however, sufficient service water flow was available from the three operating service water pumps. This failure did not aggravate this event. The site declared an Unusual Event (UE) at 1317 EDT for loss of power to emergency buses. At 0011 EDT on 10 09'16, the UE was terminated. Once the power grid was stable, plant personnel commenced restoration of offsite power to allow shutdown of the the EDGs. During this evolution, at approximately 2323 EDT on 10 . 08 2016, an automatic actuation of the 'B' auxiliary feedwater (AFW) pump occurred due to improper breaker coordination that satisfied the autostart logic for the AFW system. The apparent cause of the voltage transient in the HBRSEP2 switchyard is a failed fault detection relay, which prevented the grid fault from being immediately isolated. The failed relay has been replaced. |
| 05000261/LER-2016-004 | 24 August 2016 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | At 1338 hours EDT On 08 24 2016, with the Unit in Mode 1 at 100 percent power, H. B. Robinson Steam Electric Plant, Unit 2 (RNP), experienced an automatic turbine trip followed by an automatic reactor trip during the performance of a visual inspection of the Main Turbine Trip Block. Following the reactor trip, the Auxiliary Feedwater System actuated as expected on low steam generator level. There were no other equipment performance issues. The cause of this event is attributed to the absence of a standard that would govern risk assessment and control of work activities around trip sensitive equipment. The perceived low significance nature of the risk in conjunction with the evolution comprising only a visual inspection and the procedural silence with respect to data gathering activities in the proximity of trip sensitive equipment, taken together gave rise to the unintended turbine trip, and a subsequent reactor trip. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) due to the event resulting in an automatic actuation of the Reactor Protection System (RPS) and Auxiliary Feedwater System (AFW). |
| 05000261/LER-2016-003 | 11 August 2016 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(ii) | The prevention of flood intrusion on plant grade at H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP), is achieved through lake level control that is performed by two electrically-operated Tainter Gates under normal operation. On 06/062016 and 07 07 2016 with RNP in Mode 1 at 100 percent power, Tainter Gates B and A were respectively identified as degraded due to the failure to meet their functionality test requirements (excessive binding of the the lift gate chains due to corrosion). RNP License Renewal program requires full travel test for the radial arm Tainter Gates to demonstrate functionality of the gates and identify any issues with operation. The Tainter Gates have been repaired, tested and rendered compliant with their original licensing basis function. On August 11, 2016 (ENS #52174), it was determined this event or condition placed RNP in an unanalyzed condition that significantly degrades plant safety. No actual Probable Maximum Precipitation Design Basis Event (PMP DBE) occurred during the period of time the Tainter Gates were in a degraded condition. Thus, this LER only considers the loss of the Emergency Diesel Generators (EDGs) and the resulting Extended Loss of Alternating Power (ELAP), if a PMP DBE had occurred. This report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(A) and 10 CFR 50.73(a)(2)(v)(D), "Prevention of Fulfillment of Safety Function," and 10 CF CFR 50.73(a)(2)(ii) (B), "Degraded or Unanalyzed Condition. |
| 05000261/LER-2016-002 | 13 April 2016 13 June 2016 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | At 1430 hours EST on 4'13'2016 with H. B. Robinson Steam Electric Plant, Unit No. 2 in Mode 1 at 100 percent power, it was determined that the source document for the mass and energy release parameters used to determine the containment pressure and temperature response to a main steam line break inside containment does not adequately account for all possible single active failure scenarios in the steam or feedwater line isolation provisions. The source document addresses the active failure of the main feedwater regulating valves to close in the faulted steam generator feedline, but not the failure of a feedwater regulating bypass valve to close in that feedline. An active failure of a feedwater regulating bypass valve whereby the valve fails to close will increase the secondary mass available for release to the containment structure. This can result in higher peak containment pressure that could challenge the containment design pressure. This condition is only a concern when the feedwater regulating bypass valves are in the open position in Modes 1, 2, or 3, and they fail to close on an engineered safeguards actuation signal. Administrative controls and corrective actions have been implemented to maintain design control of the feedwater regulating bypass in these modes of operation. |
| 05000261/LER-2016-001 | 19 January 2016 21 March 2016 | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) | At 1540 hours EST on 01/19/2016 with H. B. Robinson Steam Electric Plant, Unit No. 2, in Mode 1 at 100 percent power, Motor- Driven Auxiliary Feedwater (MDAFW) pump "B" failed its surveillance test due to flow switch FSL-1633B failure to indicate service water flow to the Oil Cooler and Packing Channel Cooler. It was discovered that flow was blocked due to stem and disc separation on the MDAFW pump "A" and "B" cooling flow return isolation valve, SW-115. This blockage rendered both MDAFW trains inoperable for a period longer than allowed by plant Technical Specifications (TS). TS 3.7.4, Auxiliary Feedwater (AFW) System, requires four AFW flow paths and three AFW pumps be operable in Modes 1, 2, and 3, and Mode 4 when steam generator is used for heat removal, allowing 24 hours to restore one of two inoperable MDAFW pumps to operable status. Since two MDAFW pumps were inoperable due to common cause for approximately 52 hours, this condition is reportable as a condition prohibited by TS under 10 CFR 50.73(a)(2)(i) (B), and reportable under 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability of Independent Trains or Channels. The total duration of both trains of MDAFW being unavailable was limited to approximately 3 hours. An investigation concluded that the failed valve (SW-115) was likely installed in poor condition, without documentation, and outside of design specifications prior to 1978. The valve has been replaced with a valve meeting current design specifications. The health and safety of the public was not jeopardized as a result of this condition. |
| 05000261/LER-2015-003 | 19 January 2015 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | At 0920 hours EST on 01/19/2015 with H. B. Robinson Steam Electric Plant, Unit No. 2 in Mode 1 at 100 percent power, the plant entered Technical Specifications (TS) Surveillance Requirement (SR) 3.0.3 due to the determination that a previously conducted test (06/18/2014) of Station Battery Charger A-1 was found to have failed due to an out of tolerance (OOT) with the measurement and test equipment (M&TE). An immediate determination of operability was conducted and it was determined that the Station Battery Charger A-1 was operable. On 02/17/2015, a review of reportability for the Station Battery Charger A-1 M&TE issue was approved and concluded that the condition was not reportable since Station Battery Charger A-1 was believed operable from the time of the test conducted on 06/18/2014 even with the M&TE OOT. On 02/24/2015, an investigation concluded that the entry into SR 3.0.3 was not appropriate, but no review of the change in reportability was conducted. On 04/14/2015, it was determined that the condition was reportable. This condition was present for a time greater than allowed by TS and is reportable as a condition prohibited by TS under 10 CFR 50.73(a)(2)(i)(B). Compliance with TS 3.8.4 Condition A was restored on 01/19/2015 when Station Battery Charger A-1 was removed from service and Station Battery Charger A was placed in service. |
| 05000261/LER-2015-002 | 18 March 2015 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | At 1703 hours EST on 03/18/2015 with H. B. Robinson Steam Electric Plant , Unit No. 2 in Mode 1 at 100 percent power, it was determined that the limit switches that provide indication in the control room as to the position of the pressurizer power operated relief valves (PORVs) may be at the end of their service life. Further investigation revealed that the qualified life requirements of 10 CFR 50.49 and Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, for the limit switches was incorrect in station environmental qualification documentation. The qualified life of the limit switches expired on 2/23 2007. The normal operation of the PORV limit switches in non-accident conditions is not limited, and they are considered operable in all modes of operation except during a design basis accident. However, the limit switches are considered inoperable for the purposes of meeting Technical Specification (TS) 3.3.3, Post Accident Monitoring Instrumentation. Therefore, this circumstance is reportable as a condition prohibited by TS under 10 CFR 50.73(a)(2)(i)(B). Standing Instruction 15-005, effective on March 20, 2015, outlines the alternate method for continuous and post-accident monitoring of pressurizer PORV position until the limit switches are returned to service. Work orders have been generated to replace the limit switches during the upcoming refueling outage, currently scheduled to begin on May 12, 2015. |
| 05000261/LER-2015-001 | 28 January 2015 | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | At 1957 hours EST on 01/28/2015 with H. B. Robinson Steam Electric Plant, Unit No. 2 in Mode 1 at 100 percent power, the plant entered Improved Technical Specifications (ITS) Limiting Condition for Operation (LCO) 3.0.3 due to inoperability of both trains of the Reactor Protection System (RPS). It was discovered that modification work performed during the Fall 2013 refueling outage inadvertently connected in parallel both safety trains of the RPS and both trains of the DC Electrical Distribution System (DC-EDS). This parallel connection rendered both of these systems inoperable because the required independence and redundancy of systems was eliminated. This condition was present for a time greater than allowed by technical specifications (TS) and is reportable as a condition prohibited by TS under 10 CFR 50.73(a)(2)(i)(B). This condition also rendered inoperable one train of multiple systems designed to mitigate the consequences of an accident and is reportable under 10 CFR 50.73(a)(2)(vii). Immediate actions taken to restore compliance with regulations included completion of an emergency Work Order that restored the independence of safety trains of both systems and permitted exit of ITS LCO 3.0.3 at 2048 on 01/28/2015. The condition was entered into the site corrective action program and a root cause investigation is in progress. |
| 05000261/LER-2014-002 | 26 June 2014 | On 11/26/2014 the Robinson Nuclear Regulatory Commission Resident notified plant staff that the practice of crediting the plant computer (Emergency Response Facility Information System or ERFIS) in lieu of Rod Position Indication (RPI), and not entering the required actions of Technical Specification (TS) 3.1.7 had been reviewed and concluded that this practice was not allowed per the TS. One occurrence was found where the required actions for TS 3.1.7 were not entered due to crediting ERFIS. On 6/26/2014, Operators identified a discrepancy between RPI indicator and ERFIS indication for a single control rod. Licensed Operators determined that entry into LCO 3.1.7 was not required and ERFIS was documented as acceptable for RPI system operability. All ERFIS RPI indications remained normal and within limits. Troubleshooting of the RPI on 06/30/2015 revealed that the signal conditioner to the RPI display had failed and the ERFIS display for the control rod in question was accurate, showing that the control rod was within limits; therefore, there were no safety impacts from this event. Actions were to suspend this practice and issue Standing Instruction 14-023. Operations Surveillance Test procedures were revised to remove language that supported operator actions to credit ERFIS data. The condition was entered into the site's corrective action program. | |
| 05000261/LER-2014-001 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | At 2234 hours EST on 1/9/2014, with the Unit in Mode 1 at 100% power, H. B. Robinson Steam Electric Plant, Unit 2 experienced an automatic reactor trip/turbine trip during the performance of Steam Generator (SG) Water Level Protection Channel testing. The 'EV reactor trip breaker opened as a result of its 2/3 SG Lo-Lo Level input logic being satisfied. This occurred when one channel contact was "open" due to foreign material lodged between the contact faces and the second channel contact was opened during channel testing. The opening of the 'EV reactor trip breaker resulted in a turbine trip followed by a reactor trip. Auxiliary Feedwater automatically started as expected. There were no other equipment performance issues. The cause of this event was degradation of passive components (wire labels) in the Reactor Protection System (RPS) relay rack. Degraded wire labels were the source of the foreign material which became lodged in an RPS relay contact creating an undetected half- trip condition. The foreign material was removed from the RPS relay contact. Both trains of RPS were tested to verify proper functioning of each RPS relay, and both trains of RPS relay racks were inspected to confirm no foreign material was present which could affect proper operation of the RPS relays. | |
| 05000261/LER-2013-003 | At 1801 hours EST on 11/5/2013, with the Unit in Mode 1 at 19% power, H. B. Robinson Steam Electric Plant, Unit No. 2 (1-IBRSEP2) experienced an automatic reactor trip while operators were transferring loads from the Startup Transformer to the Unit Auxiliary Transformer (UAT), as part of activities associated with coming on line and increasing power following Refueling Outage'28. During the coordinated breaker operation, the reactor tripped when an anomaly occurred that resulted in the actuation of two undervoltage emergency bus as a result of the undervoltage transient. The event was reported as a 4 hour Non-Emergency report per 10 CFR 50.72 (b)(2)(iv)(B) due to the valid Reactor Protection System Actuation, and as an 8 hour Non-Emergency report per 10 CFR 50.72(b)(3)(iv) (A) due to the valid actuation of Auxiliary Feed Water and EDG auto-start and subsequent starting of required undervoltage loads, save `A' Service Water Pump. The investigation into the cause of this event determined that advanced aging/fatigue of the phenolic operating rods of the UAT breaker (52/7) caused the failure of the 'EV phase operating rod, which prevented closure of the 'EY phase of the breaker. The cause of the failure of the 'A' Service Water Pump to sequence onto the E-1 bus during the blackout sequence was attributed to a loose wire termination in the Emergency Control Station (ECS). Immediate corrective actions consisted of assessment of the switchyard and E-1 Bus, inspection of Breaker 52/7, and securing the loose wire termination in the ECS. This event did not impact the health and safety of the public. | ||
| 05000261/LER-2013-002 | At 0041 hours EST on 11/05/2013, with the Unit in Mode 2 and startup low-power physics testing in progress, there was an inadvertent automatic actuation of the Auxiliary Feedwater System due to an 'A' Main Feed Pump trip. While placing the Condensate Polishers in service, a secondary-side perturbation occurred resulting in the loss of the running 'A' Main Feedwater Pump on low suction pressure coincident with low flow. There was no plant damage or personnel injury as a result of this event. An 8-hour, non-emergency notification was made to the NRC per 10 CFR 50.72(b)(3)(iv)(A) due to the valid actuation of the Auxiliary Feedwater System (EN# 49502). The root cause evaluation has concluded that the cause of this event was an operating error by one individual involving procedure use and adherence. Corrective actions consist of completion of confidential personnel actions for the Makeup Water Treatment/Condensate Polisher Auxiliary Operator responsible for the event, and enhancement of expectations for supervisory oversight of risk significant evolutions will be employed. | ||
| 05000261/LER-2013-001 | On 10/6/2013, with the plant de-fueled and vessel head removed, it was discovered that a non-environmentally-qualified butt-splice was installed in a wire connected to the 'closed' limit switch for a containment isolation valve, which rendered the Post-Accident Monitoring (PAM) Instrumentation function - Containment Isolation Valve Position Indication - inoperable. This condition has been present for an extended period of time, and it is presumed that on multiple occasions this function was inoperable for a period of time greater than allowed by Technical Specifications (TS) 3.3.3, PAM Instrumentation Limiting Conditions for Operation. The initial investigation into the cause of this event indicates this was an isolated human performance event in which the non-licensed air-operated valve (AOV) technician failed to use proper material specified for the task per the procedure directing the task. Immediate corrective action consisted of the removal of non-environmentally-qualified splice and subsequent installation of an environmentally- qualified splice, which returned the component to operable condition. This event did not impact the health and safety of the public. | ||
| 05000261/LER-2011-002 | 26 September 2011 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | |
| 05000261/LER-2011-001 | 4 May 2011 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | |
| 05000261/LER-1986-012, Forwards LER 86-012-01,revising Level of Reactor Power from Which Unit Tripped from 42% to 13%.Operating Mode Marked N as Recommended by NUREG-1022 | 15 January 1987 | ||
| 05000261/LER-1984-003, Forwards LER 84-003-01 Containing Complete Description of Event & Current Corrective Actions Re Anomalies W/Steam Generator Snubbers While in Test Stand Under Simulated Conditions | 12 December 1984 | ||
| 05000261/LER-1983-005, Supplemental LER 83-005/01T-1:on 830424,svc Water Booster Pump B Declared Inoperable Due to Loss of Bearing Oil.Caused by Bearing Oil Slinger Positioned Near Vent Holes.Standpipes Installed on Each Vent Hole | 14 June 1984 | ||
| 05000261/LER-1983-003, Forwards LER 83-003/01T-0 | 22 April 1983 | ||
| 05000261/LER-1983-002, Forwards LER 83-002/03L-0 | 17 March 1983 | ||
| 05000261/LER-1983-001, Forwards LER 83-001/01L-0 | 21 February 1983 | ||
| 05000261/LER-1982-019, Forwards LER 82-019/01L-0 | 31 December 1982 | ||
| 05000261/LER-1982-018, Forwards LER 82-018/01L-0 | 23 December 1982 | ||
| 05000261/LER-1982-016, Forwards LER 82-016/03L-0 | 30 November 1982 | ||
| 05000261/LER-1982-015, Forwards LER 82-015/03L-0 | 17 November 1982 | ||
| 05000261/LER-1982-014, Forwards LER 82-014/03L-0 | 11 November 1982 | ||
| 05000261/LER-1982-013, Forwards LER 82-013/03L-0 | 22 October 1982 | ||
| 05000261/LER-1982-012, Forwards LER 82-012/03L-0 | 6 October 1982 | ||
| 05000261/LER-1982-011, Forwards LER 82-011/03L-0 | 15 September 1982 | ||
| 05000261/LER-1982-010, Forwards LER 82-010/01T-0 | 17 August 1982 | ||
| 05000261/LER-1982-009, Forwards LER 82-009/03L-0 | 13 August 1982 | ||
| 05000261/LER-1982-008, Forwards LER 82-008/03L-0 | 9 August 1982 | ||
| 05000261/LER-1982-007, Forwards LER 82-007/01T-0 | 31 July 1982 | ||
| 05000261/LER-1982-006, Forwards LER 82-006/01T-0 | 28 July 1982 | ||
| 05000261/LER-1982-005, Forwards LER 82-005/01T-0 | 12 July 1982 | ||
| 05000261/LER-1982-001, Forwards LER 82-001/03L-0 | 24 February 1982 | ||
| 05000261/LER-1981-031, Forwards LER 81-031/01T-0 | 15 December 1981 | ||
| 05000261/LER-1981-027, Forwards LER 81-027/03L-0 | 15 December 1981 | ||
| 05000261/LER-1981-008, Forwards LER 81-008/03L-0 | 27 March 1981 | ||
| 05000261/LER-1981-002, Updated LER 81-002/01T-2:on 810113,review of Svc Water Sys Revealed Potential Unmonitored Release Path on Svc Water Return Lines from Containment for Hydrogen Vent Header Motor Coolers.Caused by Design Error.Sys Will Be Modifie | 20 August 1982 | ||
| 05000261/LER-1981-001, Updated LER 81-001/01T-1:on 810109,turbine Runback & Automatic Rod Withdrawal Block Protection Defeated W/Power Range Channel N41 Being Out of Svc.Caused by Improper Rod Drop Analysis.Administrative Controls Revised | 17 November 1982 | ||
| 05000261/LER-1980-030, Updated LER 80-030/03L-1:on 801202 & 03,charging Pumps A,B & C Discharge Line Pressure Gauge Tap Pipe Socklets Found Leaking.Caused by Sys Design Deficiencies Allowing High Vibration.Leaks Repaired & Design Modified | 30 September 1983 | ||
| 05000261/LER-1980-027, Updated LER 80-027/01T-1:on 801122,relay RT-9 in Reactor Protection Train a Failed to re-energize During Periodic Test of Reactor Protection Logic.Caused by Shorting of Lead Wire Connection to Coil Wires.Relay Replaced | 11 October 1982 | ||
| 05000261/LER-1976-014, Updated LER 76-014:on 760805,util Notified by NRC That Peak Clad Temp Increase Presented by Westinghouse Should Be 50 F Higher.Caused by Using Different Versions of ECCS Model. New Limit in Effect 750825.W/760917 Ltr | 17 September 1976 |