Semantic search

On December 13, 2016, the Unit 1 AB Emergency Diesel Generator (EDG) developed a fuel oil leak from a fuel injector pump Delivery Valve Holder (DVH) during a maintenance run of the diesel. On December 21, 2016, with Unit 1 in Mode 1 at 100 percent power and Unit 2 in Mode 4 during a refueling outage, it was determined that the failed DVH on the Unit 1 AB EDG was due to a design and manufacturing issue. Subsequently the Unit 1 CD EDG, Unit 2 AB EDG, and Unit 2 CD EDG were conservatively declared inoperable due to multiple affected diesel fuel pump DVHs being installed on each EDG.

The Root Cause was determined to be due to insufficient Corrective Action Program oversight by Engineering to ensure product quality and issue resolution in relation to a previous EDG DVH failure in 2013. Testing is being conducted to determine the resulting impact to associated EDGs. The affected DVHs have been replaced.

A Loss of Safety Function was reported via Event Notification 52456 for Unit 2 in accordance with 10 CFR 50.72(b)(3)(v)(D). The Loss of Safety Function is required to be reported in a Licensee Event Report in accordance with "50.73(a)(2)(v)(D) Event or Condition that Could Have Prevented Fulfillment of a Safety Function.

On April 23, 2015, at 0210, Donald C. Cook Nuclear Plant Unit 2 Reactor was manually tripped from approximately 2 percent of rated thermal power during plant restart following a refueling outage. Unit 2 Reactor was manually tripped due to the inability to maintain Average Reactor Coolant System Temperature above the Technical Specification (TS) required minimum Temperature for Criticality when two newly installed Steam Dump Valves failed open while being manually valved into service. The valves were subjected to, but not designed for, two phase flow.

The Root Cause has been determined to be that the modification process failed to identify and document all system operational vulnerabilities. The corrective action to preclude repetition is an enhancement of the Engineering Modifications procedure to require development and inclusion of a narrative to describe system operation, including key interfacing system operation.

The manual Reactor Protection System (RPS) actuation was reported via Event Notification 51004 in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A), and 10 CFR 50.72(b)(2)(i). The valid RPS actuation and the completion of the plant shutdown required by TS are reportable as a Licensee Event Report in accordance with (- 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(2)(i)(A) respectively.

This supplemental LER is being issued to update the causal and corrective action statements. At 0400 hours on March 5, 2003, Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.1.2, Action "a" was entered to perform routine maintenance on the Unit 2 West Motor Driven Auxiliary Feedwater (MDAFW) pump. During post-maintenance testing, a loud 'buzzing" noise was heard emanating from the vicinity of the pump motor. A supporting/refuting evaluation was performed, which eliminated the pump, breaker, or system alignment as the possible source of the noise. At this point, the decision was made to replace the suspect motor. Initially, Indiana Michigan Power (I&M) had expected to complete the replacement of the West MDAFW pump motor within the allowed outage times specified in TS 3.7.1.2. However, unanticipated delays prevented the completion of this activity within the allowed outage time.

Therefore, I&M requested, and was granted, enforcement discretion on March 8, 2003. .

The apparent cause of the Unit 2 West MDAFW pump motor noise was loose air baffles due to mounting hole deformation as determined by the vendor's as found condition analysis. The cause for the failure to complete the motor replacement within the allowed outage time was inconsistent maintenance work practices and inadequate interface requirements that resulted In a lack of command and control process for short term LCO activities. Corrective actions to prevent recurrence included the establishment of administrative guidance to ensure the appropriate techniques and tools are used when performing work on critical components, and the development and satisfactory completion of coupling installation training. Also, going forward, KM established a team to respond to equipment issues that challenge short duration allowed outage time.

I&M determined that no net Increase in risk was associated with extending the TS 72-hour allowed outage time by an additional 36 hours to restore the West MDAFW pump to an operable status. Although the proposed action deviated from a requirement In TS 3.7.1.2, it did not affect any safety limits, setpoints in the TS, or other operational parameters, nor did it affect any margins assumed In the accident analyses. In addition, the redundant Unit 2 East MDAFW pump and TDAFW pump continued to be operable to perform their required design function. The Unit 2 West MDAFW pump motor was replaced with a spare motor and the pump was declared operable at 0246 hours on March 9, 2003.

At 0045 hours on July 22, 2002, Unit 2 tripped due to low condenser vacuum. At the time of the trip, operators were performing a flush of the main condensers in preparation for the upcoming biocide treatment of the circulating water system. In accordance with 10 CFR 50.72 (b)(2)(iv)(B), a four-hour ENS notification (Event # 39081) was made to the NRC on July 22, 2002, at 0233 hours for an event or condition that resulted in an actuation of the reactor protection system (RPS) when the reactor is critical. As such, this LER is being submitted in accordance with the requirements of 10 CFR 50.73 (a)(2)(Iv)(A) for a condition or event that resulted in an automatic actuation of the RPS system.

The apparent cause of the event is a previously unrecognized steam side heat transfer anomaly in "C" main condenser that has resulted in "C" North water box removing significantly more heat than in "C" South water box. Corrective actions to prevent recurrence of this event will be determined by an equipment root cause investigation to be performed after the steam side of the "C" main condenser has been inspected during the next Unit 2 outage of sufficient duration. Additional corrective actions, including preventive actions, may be developed based on the results of the investigation. If significant changes are identified as a result of completion of the root cause investigation, an update to this LER will be submitted.

As a compensatory action, the main condenser water box flushing and isolation procedure has been revised to provide improved guidance for performing these activities to minimize the possibility of a unit trip The safety significance of this event was minimal. Plant systems responded per design with no significant anomalies noted. Plant procedures and operator training provided sufficient direction for control room personnel to shutdown the plant and maintain it in a safe shutdown condition. There was no impact on the health and safety of the public as a result of this event

At 00:01 hours on 01/19/02, in preparation for a Unit 2 refueling outage, Operations shift personnel initiated a planned manual reactor trip of Unit 2 from 22t power per Procedure 02-OHP-4021-001-003, Revision 15, T "Power Reduction." T Shortly thereafter, an automatic start of the turbine driven auxiliary feedwater pump (TDAFP) occurred as a result of a valid low-low level indication in the steam generators. T The automatic start of the TDAFP was determined to be an "unanticipated" engineered safety feature (ESF) actuation.

Steam generator levels rapidly recovered. T Operators secured the"TDAFP and throttled the flows from the motor driven auxiliary-feedwater pumps in accordance with plant procedures for reactor trip response and recovery. T Reactor coolant system cooldown and depressurization proceeded normally. T During the trip, pressurizer level shrank lower than procedurally anticipated, resulting in a reactor coolant system 'letdown isolation.

At 07:56 on 01/19/02, the Shift Manager made an eight hour, non-emergency notification to the NRC (EN# 38640) per 10 CFR 50.72(b)(3)(iv)(A) for an unanticipated ESF actuation. T The cause of this event was inadequate procedural guidance. T Corrective actions included revision of the applicable procedures to include a reduction in the planned power level trip point to reduce the potential for automatic start of the TDAFP.

On February 12, 2002, Technical Specification (TS) 3.9.4.c was violated when the 100 PSI Control Air to Containment Control Air Header #2 Train 'B' Containment Isolation Valve, 2-XCR-101(EllS:LKSHV), was stroked open during core alterations with open test connections on both sides of the valve, one inside containment and one outside. This configuration provided direct access from the containment atmosphere to the outside atmosphere. The apparent cause was determined to be ineffective procedural control (02-OHP-4030.STP.041). The preferred method for establishing refueling integrity for 2-CPN-74 in the procedure does not consider the possibility that the control air ring headers may be cross-tied. A contributing cause was that the installation of containment control air header cross-tie jumpers was not documented in the Proceduralized Temporary Modification Log on January 20, 2002.

Upon discovery that the preferred method used to maintain refueling integrity for 2-CPN-74 was ineffective, an alternate method was established and valve 2-XCR-101 was visually confirmed to be intact and closed. The Temporary Modification Log Index was reviewed to verify that other Temporary Modifications were logged. The containment control air cross-tie jumpers were documented in the log at that time.

The refueling integrity procedures will be revised to ensure that the methods used to establish refueling integrity for Penetrations CPN-74 and CPN-29 recognize that the Containment Control Air headers may be cross-tied. Unit 1 and Unit 2 Type B and C Leak Rate Testing procedures will be revised to ensure that steps are added for documenting the placement of the containment control air header cross tie jumpers in the Proceduralized Temporary Modification Log.

On October 7, 2001, a reactor trip occurred at 8 percent reactor power. The trip was the result of a loss of rod control system voltage. The cause of the loss of rod control system voltage was an open resistor at the input to the north control rod drive motor generator (CRD-MG) voltage regulator. The open resistor caused a low voltage transient when the north CRD-MG field collapsed. A protective auxiliary relay removed power from the south CRD-MG voltage regulator resulting in a loss of rod control system voltage. The loss of voltage caused all control rods to rapidly insert, thereby, initiating a Power Range, Neutron Flux, High Negative Rate trip from the reactor protection system (RPS). The RPS actuation was initiated by actual plant conditions that satisfied the requirements for the initiation of the trip, and was, therefore, a valid RPS actuation.

The failed resistor was determined to be the result of a random component failure. The resistor was original equipment and inspection revealed no evidence of overheating or manufacturing defect. The defective resistor was replaced and, as a precaution, the remaining series resistor in the north CRD-MG along with the identical resistors in the south CRD-MG were replaced.

This event was reported in accordance with 10 CFR 50.72(b)(1)(iv)(B).

This LER supplement is being submitted to include revised information related to the completed root cause evaluation.

This LER revision replaces the previous LER in its entirety. On January 23, 2001, during the removal of plant equipment from the Unit 2 lower containment personnel airlock, the airlock doors' interlock failed. This allowed the inadvertent opening of both the inner and outer lower containment airlock doors at the same time for approximately 5 seconds.

Technical Specification 3.6.1.3 requires both containment airlock doors to be closed; except during normal transit entry and exit through containment, then at least one airlock door shall be closed. Because both lower containment airlock doors were open at the same time, an 8-hour ENS notification was made to the NRC in accordance with 10 CFR 50.72 (b)(3)(v)(C), for a condition or event that could have prevented the fulfillment of the safety function of a system needed to control the release of radioactive material.

The root cause for the containment airlock door interlock failure was the interlock mechanism slipping out of adjustment.

The specific failure involved a gradual loosening of the setscrews that hold the interlock gears in place.

Both the inner and outer lower containment airlock doors were immediately closed to restore containment integrity. The airlock door interlock was repaired and satisfactorily tested. Preventive maintenance (PM) activities for the airlock doors were evaluated and the root cause recommendations and vendor recommendations have been incorporated into the appropriate procedures. A detailed design analysis is presently being performed and is being tracked in accordance with the site Corrective Action Program. The appropriate procedure has been revised to include directions on the operation of the containment airlock doors and the consequences of improper airlock door configuration, including radiological and industrial safety concerns. This condition is not considered to be safety significant due to the extremely low probability of a Loss of Coolant Accident or Main Steam Line Break occurring during the 5-second time interval in which both airlock doors were open. A review of plant events during the past three years did not identify any conditions in which the containment doors were opened simultaneously. Therefore, this is considered an isolated event.

FORM 366 (7-2001) �