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On 11/28/13, at 0258 hours, the Control room received a Unit 2 interlock door alarm. Operations sent an equipment operator to investigate; and found a Radiation Protection Technician (RPT) and laborers in the area of the alarming interlock removing lead blankets.

entered into Technical Specifications 3.6.4.1 Condition A. The doors were closed and the Technical Specifications Condition was exited.

The cause of this event was determined to be a failure to recognize a hazard while proceeding in the face of uncertainty.

Specifically, the RPT did not recognize that there was a change in plant conditions (i.e., the restoration of the X-Area as part of the Secondary Containment boundary) and decided that it was permissible to override the door interlocks by using an emergency manual push-button. As a result of this event, Radiation Protection will be reviewing the proper use of human performance tools and will be revising procedures to enhance the notifications made to the Radiation Protection Department.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(C), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

klOr Cr101111 ncc tni and e APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/3112e Reported lessons learned are incorporated Into the licensing process and fed back to industry.

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On November 14, 2013, with Unit 2 shutdown for refueling outage D2R23, both the 2-0220-58B Feed Water Inboard Check Valve and the 2-0220-62B Feed Water Outboard Check Valve failed Local Leak Rate Testing (LLRT) acceptance criteria. Specifically, the "as-found" leak rate for both valves was above the administrative acceptance criteria of 45.0 scfh, and based upon the leakage rate observed, leakage was also determined to exceed the limits for primary containment leakage as specified in TS 5.5.12.c.

During testing, the in-series Feed Water "B" Loop Containment Isolation valve volumes could not be pressurized with full flow service air through the test tap configuration. The injected air immediately passed through the check valve and out the test vent tap, indicating that the valve disc was not fully seated. The cause of this event was determined to be that best practices, for installation and testing of the valves in order to maximize reliability, had not been implemented. The station has subsequently developed and implemented valve maintenance and testing best practices. Additionally, the station is evaluating long term actions to enhance the design and testing of the Feed Water check valves.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(A) and 10 CFR 50.73(a)(2)(v)(C).

On 6/28/2013, from 07:49:07 to 07:49:14, control room operators received a control room alarm indicating that the unit 2/3 emergency diesel generator (EDG) interlock doors were open simultaneously. Individuals were able to enter the interlock, from two directions, at the same time. With two interlock doors open simultaneously, Technical Specifications 3.6.4.1, Surveillance Requirement 3.6.4.1.2, was not met. With the Surveillance Requirement not met, Secondary Containment was declared inoperable, and entry into Technical Specifications 3.6.4.1 Condition A was made. The doors were immediately closed and the Technical Specifications Condition was exited. Based upon the investigation performed, the failure mechanism for this event was determined to be an isolated and rare premature failure of the latch bolt monitor.

Based upon the short duration of the secondary containment doors being opened simultaneously and that the Secondary Containment differential pressure remained negative during the course of this event, this event is of low safety significance.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(C), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

On March 15, 2009, at approximately 2025 hours (CDT), with Unit 2 at approximately 100 percent power, Dresden Nuclear Power Station Operations personnel discovered during a maintenance activity that motor operated valve 2-2301-6, Unit 2 High Pressure Coolant Injection Suction Valve, would not close. The valve is a normally open valve and is required to close during the transfer of High Pressure Coolant Injection System pump suction from the Condensate Storage Tanks to the Torus to prevent high Torus water level. Troubleshooting and diagnostic testing identified that the valve's failure to close was attributed to valve internal binding.

The Root Cause of this event was that when the valve was in the full open position, the center of gravity of the solid gate valve disc extends past the end of the in-body valve guides, which caused the disc to tip and bind with the in-body guides.

A contributing cause to this event is increased stem friction. The higher than normal thrust associated with the binding caused a loss of stem lubricant through extrusion, which increased the friction at the stem and stem nut interface.

The in-body valve guides were lengthened into bonnet to ensure the valve disc's center of gravity is supported by guides when the valve is in the Open position. Preventative maintenance activities for motor operated valve stem lubrication were revised, as needed, to ensure that the frequency does not exceed two-years for the motor operated valve program valves with horizontal disc/stem orientation that exhibit internal binding.

The safety significance of the event is minimal. An evaluation concluded that HPCI would have operated long enough to fulfill its safety function. A review of the actual Torus water levels during this time frame identified that the Torus would not have exceed its structural design values during a postulated accident. Therefore, the consequences of this event had minimal impact on the health and safety of the public and reactor safety.

On 10/03/2009 at approximately1735 hours while at full power, Unit 3 experienced an automatic reactor scram and Group I primary containment isolation signal. Due to the Group I isolation signal, the inboard and outboard main steam isolation valves closed as designed. Prior to the reactor scram, operators were restoring the reactor water clean-up (RWCU) system per the applicable station procedure.

The Probable Cause for the Group I isolation signal and reactor scram is attributed to a hydraulic pressure transient when restarting the RWCU system due to a latent procedure deficiency.

The procedure for restarting the RWCU has been revised.

During the reactor scram, the shared Unit 2/3 emergency diesel generator automatically started when auxiliary power transferred from the main to the reserve power source.

The safety significance of this event is minimal as plant response and operator actions were consistent with the protection of public health and safety and personnel safety.

On April 23, 2008, at approximately 2300 hours (CDT), with both Units 2 and 3 operating at approximately 100 percent power, the Control Room Emergency Ventilation Air Conditioning System was in operation to perform a post maintenance test. During the activity, the Control Room Emergency Ventilation Air Conditioning System experienced excessive vibration. Technical Specification 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System," was entered. The system was restored to operable status on May 1, 2008. This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(D), "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident," as the Control Room Emergency Ventilation Air Conditioning System is a single train system.

The apparent cause is attributed to the lack of lubrication due to the oil in the compressor being displaced by refrigerant. Liquid refrigerant accumulates in the compressor due to liquid floodback during operation and flooded starts. Following the completion of the actions associated with the equipment apparent cause analysis, Plant Engineering will evaluate system operation to determine the effectiveness of the actions taken. Based on this evaluation, if it is determined that system reliability has not been improved, further actions will be developed and implement, as appropriate.

On April 6, 2006, at approximately 1039 hours (CDT), with Unit 2 at approximately 97 percent power, Dresden Nuclear Power Station Instrument Maintenance technicians were installing a new High Pressure Coolant Injection System temperature recorder 2-2340-9 when an energized recorder electrical lead was accidentally shorted to a recorder mounting bracket screw. A consequence of this electrical short was the transfer of the Essential Service System to its emergency power supply from Motor Control Center 28-2 and the trip of Essential Service System circuit 16. This resulted in the High Pressure Coolant Injection System being declared inoperable due to a loss of its automatic function from the loss of power to its flow controller and signal converter. However, the system was available due to control room personnel being able to manually control system operation.

The apparent cause of this event was human performance associated with the technicians' lack of attention to detail in manipulating tools and failure to self-check. Corrective actions addressed the human performance issues through discussions of this event with the IM technicians during department training sessions and the evaluation the technicians during out-of-the-box training for proper use of human performance tools. These corrective actions reinforced the need for self-checking and the use of tape on adjacent metal components when working with energized electrical leads.

On February 3, 2005, at 1915 hours (CST), with Unit 2 at approximately 95 percent power and Unit 3 at approximately 96 percent power, Dresden Nuclear Power Station engineering personnel confirmed that single failure vulnerabilities existed on 4160 Volt Relaying and Metering transformers on both units. The single failure vulnerabilities defeated the independence of the offsite alternating current power supplies. If this failure occurred during a Loss of Coolant Accident, the Containment Cooling Service Water system pumps may not have been able to be started within the required time. The single failure vulnerabilities were removed on February 4, 2005, at 0009 hours, and the effected plant equipment was declared operable.

The root cause of this event was determined to be an existing latent design deficiency. The corrective action to prevent reoccurrence is currently in place in the Configuration Change procedures used to install new designs at Dresden Nuclear Power Station. A preliminary engineering review for similar existing latent design deficiencies found no other similar deficiencies. A corrective action is in place for engineering to complete an extent of condition review of the alternating current, Emergency Diesel Generators, and direct current systems for latent design deficiency conditions similar to this event.

On September 29, 2004, at 1235 hours (CDT), with Unit 3 at approximately 100 percent power in Mode 1, it was discovered that 3 of the 4 Unit 3 Isolation Condenser Time Delay Relays exceeded their Allowed Value specified in Technical Specification 3.3.5.2, "Isolation Condenser (IC) System Instrumentation." The event was discovered while performing Surveillance Procedure DIS 1300-08, "Sustained High Reactor Pressure Time Delay Relay Calibration." This event identified that two independent channels of the time delay portion of the Unit 3 Isolation Condenser Initiation Logic were inoperable and is being reported under 10 CFR 50.73(a)(2)(vii), "), "Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system." Additionally, this event is being reported under 10 CFR 50.73(a)(2)(i)(B), "Operation or condition prohibited by Technical Specifications." The Isolation Condenser Time Delay Relays were previously calibrated on June 16, 2002 by using a stopwatch.

The root cause of the event was due to an ineffective extent of condition review for a corrective action associated with a 1996 event in which stopwatches were determined to not be sensitive enough for calibration checks on components with limited margin. The corrective action to prevent recurrence was to use a strip chart recorder for calibrating the Isolation Condenser Time Delay Relays.