|Report date||Site||Event description|
|05000261/LER-2017-002||14 December 2017||Robinson|
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||2 June 2017||Robinson|
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||22 February 2017||Robinson|
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 October 2016||Robinson|
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||10 October 2016||Robinson|
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 June 2016||Robinson||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||21 March 2016||Robinson|
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-2015-002||17 November 2015||Robinson|
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||30 March 2015||Robinson|
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 January 2015||Robinson|
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-2013-002||19 June 2014||Robinson|
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-2011-002||15 November 2011||Robinson|
|05000261/LER-2011-001||5 July 2011||Robinson|