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 Report dateSiteEvent description
05000254/LER-2017-0045 January 2018Quad Cities

During performance of the High Pressure Coolant Injection (HPCI) Pump Operability Test, the Unit 1 HPCI turbine did not trip when the Remote HPCI Turbine Trip pushbutton was depressed. Operations shut down HPCI by isolating the steam supply valves to trip the HPCI turbine. The Unit 1 HPCI system was declared inoperable, but remained available. The cause was determined to be accumulation of wear debris within the HPCI turbine stop valve oil resetting solenoid valve causing the valve to stick in the energized position. This wear debris was a result of a manufacturing deficiency.

The immediate corrective action was to replace the turbine stop valve oil resetting solenoid valve. Follow-up corrective action is to evaluate the preventative maintenance frequency for the HPCI turbine stop valve oil resetting solenoid valve.

The safety impact of this condition was minimal. The HPCI system was still available to function, despite the issue with the HPCI Turbine Stop Valve oil resetting solenoid. The event is being reported because HPCI is a single train system and the loss of HPCI could potentially impact the plant's ability to mitigate the consequences of an accident.

05000254/LER-2017-00317 November 2017Quad Cities

On 09/21/2017 at 1550, Operations started Control Room Emergency Ventilation (CREV) Air Conditioning (AC) for Tracer Gas Testing. Per Operations instructions, Mechanical Maintenance went to inspect the Control Room Emergency Ventilation system for refrigerant leaks before Tracer Gas Testing was started. Mechanical Maintenance reported a refrigerant leak on the discharge piping of the compressor, right above the inlet to the condenser. The leak was at the expansion joint of a fitting. The safety significance of this event was minimal.

The cause of the refrigerant leak on the Control Room Emergency Ventilation compressor discharge pipe fitting into the condenser was due to high cycle fatigue.

The corrective action was to replace the failed Control Room Emergency Ventilation compressor discharge pipe fitting.

The CREV AC system is a single train system. Given the impact on the CREV AC system, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of 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.

05000265/LER-2017-00114 July 2017Quad Cities

On May 15, 2017, at 19:18 hours, station Operations personnel were performing a High Pressure Coolant Injection (HPCI) Pump Operability Test which ensures the HPCI Minimum Flow Valve opens as pump flow decreases. When the HPCI Turbine was tripped, the Minimum Flow Valve did not open as expected when system flow was reduced to the low flow setpoint. Operators took steps to open the valve manually, but upon release of the control switch, the valve returned to the closed position.

The valve was then left in the closed position.

The HPCI system was declared inoperable and Technical Specification 3.5.1 Condition G was entered.

The cause of the Minimum Flow Valve failing to open was attributed to the HPCI Pump Discharge Flow Indicating Switch, specifically, intermittent failure of the high side micro switch caused by residual material from the manufacturing process.

The Flow Indicating Switch, which had been installed for three months, was replaced and the HPCI Pump Operability Test was successfully re-performed. The failed switch was then sent to Exelon's Power Labs for failure analysis.

The safety significance of this event was minimal. Given the impact on the HPCI system, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of 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. The HPCI system is a single train system and the loss of HPCI could impact the plant's ability to mitigate the consequences of an accident.

05000254/LER-2017-00226 May 2017Quad Cities

On March 27, 2017, during refueling outage Q1R24 at 0840 hours, the Unit 1 Main Steam Isolation Valve (MSIV) as-found closure time test results indicated that four MSIVs failed to close within the required Technical Specification (TS) time of less than or equal to five seconds. The safety significance of this event was minimal.

The cause of three of the slow closure timing events was due to a less than optimal replacement frequency for the MSIV actuators, and the cause of the fourth slow closure timing event was due to a less than optimal replacement frequency for the MSIV springs.

Corrective actions included replacing the 1-0203-1B and 1-0203-1D MSIV actuators, replacing the springs on the 1-0203-2C MSIV, readjusting the as-left closure times on all four MSIVs and satisfactory retesting prior to startup. Follow-up corrective actions include replacing the 1-0203-2D MSIV actuator and changing the preventive maintenance frequency and description.

Since the MSIV slow closure times were due to degradation from less than optimal replacement frequencies, it is likely the degradation occurred over time since the last successful refueling outage testing and during power operations when the required TS 3.6.1.3, Primary Containment Isolation Valves was applicable. Therefore, this condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), which requires reporting of any operation or condition that was prohibited by the plant's TS.

05000254/LER-2017-00122 March 2017Quad Cities

On January 24, 2017, at 10:00 hours, Operations was notified that both doors in the secondary containment interlock on the 595 foot elevation between the Reactor Building (RB) and the Unit 2 Reactor Feed Pump room (located inside the Turbine Building (TB)) were opened simultaneously for approximately 3 seconds. The failure of this interlock caused a loss of secondary containment per Technical Specification (TS) 3.6.4.1, Condition A. The doors were immediately reclosed, and the secondary containment boundary was immediately reestablished. Operators verified the RB (secondary containment) differential pressure was maintained operable at greater than 0.10 inches of water vacuum.

Secondary containment remained available and functional during the event since the secondary containment interlock was immediately restored by closing the doors and since the RB differential pressure was maintained during the event. The RB is a common volume to both Units 1 and 2, and an interlock failure can impact the secondary containment for both units.

The cause of the interlock failure was due to a dirty contact that caused the interlock relay to stick. This allowed the second door to open before the first door was secured. Corrective actions included inspecting and cleaning of the interlock relay contacts. Steps will be added to the Preventative Maintenance Work Orders to perform a visual inspection and cleaning of interlock relays.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of 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.

05000265/LER-2016-00325 July 2016Quad Cities

result, the Station determined that a drywell entry was required to investigate the condition, make any needed repairs, and refill the oil reservoir. NRC provided that Station personnel did not comply with Technical Specifications (TS) 3.6.2.5 (DW to Suppression.

Chamber DP) and 3.6.3.1 (Primary Containment 02 concentration) since while in MODE 1, at the end of the 32 hour Completion Time (24 hour Action A, plus the 8 hour Action B) during the actual plant evolutions for power ascension, these Required Actions were not met because the associated Applicability for each TS were not met since the Unit remained in MODE 1.

The cause of the issue was Station personnel understanding and application of the subject TS as used in context under this infrequent plant condition, differed from the NRC's understanding and application of the subject TS. The specific difference is with the application of the term, "start-up," as used in the LCO Applicability.

Corrective actions included issuance of an Operations Standing Order, and revision of pertinent Operating procedures to ensure these Tech Specs are properly implemented.

The safety significance of this event was minimal. Given the impact on the Drywell/Suppression Chamber Differential Pressure, and Primary Containment Oxygen Concentration Technical Specifications, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of a past operation or condition which was prohibited by the plant Technical Specifications.

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as (XX).

EVENT IDENTIFICATION

Compliance Issue with the Drywell/Suppression Chamber Differential Pressure, and Primary Containment Oxygen Concentration Technical Specifications

A. CONDITION PRIOR TO EVENT

Unit: 2 Reactor Mode: 1 Event Date: May 25, 2016 Event Time: 11:10 hours Mode Name: Power Operation Power Level: 100%

B. DESCRIPTION OF EVENT

On 05/23/16 Low Level Alarm (LA) 2A Recirc (AD) Motor (MO) occurred due to a low oil level condition for the 2A recirculation pump (P) motor. As a result, the Station determined that a drywell (NH) entry was required to investigate the condition, make any needed repairs, and refill the oil reservoir (TK). NRC provided that Station personnel did not comply with TS 3.6.2.5 (DW to Suppression Chamber DP) and 3.6.3.1 (Primary Containment 02 concentration (BB)) since while in MODE 1, at the end of the 32 hour Completion Time (24 hour Action A, plus the 8 hour Action B) during the actual plant evolutions for power ascension, the Required Actions B for TS 3.6.2.5 and TS 3.6.3.1 were not met (at 1110 on 5/25/16, and 1123 on 5/25/16, respectively), because the Unit 2 LCO Applicability for establishing DW/Torus differential pressure and being fully inerted were not met since the Unit remained in MODE 1.

NRC provided that these TS were not met when the Station improperly used the LCO Applicability (a), 24 hour "clock reset" allowance to proceed above 15% power "following startup" without setting the DW/Torus differential pressure (Dp) > 1 psid, and oxygen concentration contrary to the NRC's "plain language" interpretation of this associated TS Applicability, in that "following startup" was intended to mean "following MODE 2." Furthermore, the NRC provided that the Unit did not exit the Mode of Applicability just by dropping below 15% Rated Thermal Power (RTP), since the Unit was still in MODE 1, and a total of only 32 hours was available to re-achieve DW/Torus Dp and reinert while remaining in MODE 1. While under this interpretation, the resulting available options during this drywell entry were to either: 1) re-establish DW/Torus Dp and inerting prior to reaching 15% RTP during the power ascension, or 2) to exit MODE 1 to reset the 24 hour clock (meaning to start power ascension from MODE 2). In this situation, the TS LCO Applicability is not clear, does not coincide with the Bases intent, and may be overly restrictive in that it uses the terms, "startup" and "shutdown.

The cause of the issue was Station personnel understanding and application of the subject TS as used in context under this this infrequent plant condition, differed from the NRC's understanding and application of the subject TS.

The specific difference is with the application of the term, "start-up," as used in the LCO Applicability.

The safety significance of this event was minimal. Given the impact on compliance with the Drywell/Suppression Chamber Differential Pressure, and Primary Containment Oxygen Concentration Technical Specifications, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of a past operation or condition which was prohibited by the plant Technical Specifications.

C. CAUSE OF EVENT

During the drywell entry and subsequent return to full power, the Station performed the power ascension under procedures, QCGP 3-1 (Reactor Power Operations) and QCOP 1600-20 (Nitrogen Inerting of Primary Containment Using the Vaporizer(s) and Reactor Building Ventilation System), for which under this infrequent plant condition, the context of "startup" was understood to refer to the "act of increasing reactor power," or "power ascension." Therefore, the apparent TS interpretation conflict occurred in the meaning and use of "startup," in the LCO Applicability, when during power ascension the Station proceeded above 15% RTP without resetting DW/Torus Dp and re-inerting within the 32 hour maximum allowed Completion Time.

This issue pertained to a reading of the language of the subject TS which in itself was not readily able to be consistently interpreted since the "plain language" did not match the TS Bases nor the NRC approved text of the Safety Evaluation (SE). This TS compliance interpretation issue occurred for only a 4 hour and 2 hour duration (in excess of the 32 hours total Completion Time allowed while in MODE 1), pertaining to the DW/Torus Dp and oxygen concentration, respectively.

D. SAFETY ANALYSIS

System Design TS Bases 3.6.2.5, Drywell-to-Suppression Chamber Differential Pressure Applicable Safety Analyses provides: "The purpose of maintaining the drywell at a slightly higher pressure with respect to the suppression chamber is to minimize the drywell pressure increase necessary to clear the downcomer pipes to commence condensation of steam in the suppression pool and to minimize the mass of the accelerated water leg. This reduces the hydrodynamic loads on the torus during the LOCA blowdown. The required differential pressure results in a downcomer waterleg of approximately 1 ft. Initial drywell-to-suppression chamber differential pressure affects both the dynamic pool loads on the suppression chamber and the peak drywell pressure during downcomer pipe clearing during a Design Basis Accident LOCA. Drywell-to suppression chamber differential pressure must be maintained within the specified limits so that the safety analysis remains valid.

TS Bases 3.6.3.1, Primary Containment Oxygen Concentration Applicable Safety Analyses provides: "The UFSAR, Section 6.2.5 calculations assume that the primary containment is inerted when a Design Basis Accident loss of coolant accident occurs. Thus, the hydrogen assumed to be released to the primary containment as a result of metal water reaction in the reactor core will not produce combustible gas mixtures in the primary containment. Oxygen, which is subsequently generated by radiolytic decomposition of water, will not result in the primary containment becoming de-inerted within the first 30 days following an accident.

Safety Impact TS Bases 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure LCO Applicability provides: "As long as reactor power is containment occurring within the first 24 hours following a startup or within the last 24 hours prior to a shutdown is low enough that these "windows," with the primary containment not inerted, are also justified. The 24 hour time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting." For this event, since the period of time during which reactor power was > 15% RTP while the DW to Torus Dp was 18 hours, the probability of an event that generates hydrogen or excessive loads on primary containment was low since this duration was less than 24 hours, therefore, the safety impact of this condition was minimal.

TS Bases 3.6.3.1, Primary Containment Oxygen Concentration LCO Applicability provides: "As long as reactor power is containment need not be inert. Furthermore, the probability of an event that generates hydrogen occurring within the first 24 hours of a startup, or within the last 24 hours before a shutdown, is low enough that these "windows," when the primary containment is not inerted, are also justified. The 24 hour time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting." For this event, since the period of time during which reactor power was > 15% RTP while the DW was not inerted (i.e., 02 concentration > 4%), was approximately 16 hours, the probability of an event that generates hydrogen was low since this duration was less than 24 hours, therefore, the safety impact of this condition was minimal.

Due to the language in the associated TS Bases and SE documentation for the actions that the Station took during the drywell entry and subsequent power ascension, this TS compliance interpretation issue is not a significant event/issue, since the interpreted non-compliance occurred for only a 4 hour and 2 hour duration, pertaining to the DW/Torus Dp and oxygen concentration, respectively (i.e., 4 hour/2 hour in excess of the 32 'hours total Completion Time allowed while in MODE 1). Furthermore, this event was the first known recorded occurrence of non-compliance with these TS under this interpretation. Since the condition created no consequences, the safety impact of this condition was minimal.

Risk Insights The plant Probabilistic Risk Assessment (PRA) model was reviewed with respect to this event. Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) were evaluated for impacts of oxygen concentration and DW/Torus Dp. Since the period of time during which reactor power was > 15% RTP while the DW was not inerted (i.e., oxygen concentration > 4%), was approximately 16 hours, and since the period of time during which reactor power was > 15% RTP while the DW to Torus Dp was change in risk was minimal.

In conclusion, the overall safety significance and impact on risk of this event were minimal.

E. CORRECTIVE ACTIONS

Immediate:

1. Issued an Operations Standing Order that provided clarifying information when using the subject Tech Spec for drywell entries.

Follow-up:

2. The pertinent Operating procedures will be revised to ensure the subject Tech Specs are properly implemented for drywell entries.

3. This issue will be addressed under a proposed BWROG TSTF item for a potential future Tech Spec and Bases revision.

4. Operator Training will review this issue as an OPEX item, and for incorporation into appropriate lesson plans.

F. PREVIOUS OCCURRENCES

The Station events database, LERs, and INPO Consolidated Event System (ICES) were reviewed for similar events at the Quad Cities Nuclear Power Station. This event was caused by Station personnel understanding and application of the subject Tech Specs as used in context under this infrequent plant condition, differed from the NRC understanding and application of the subject Tech Specs. The specific difference is with the application of the term, "start-up," as used in the LCO Applicability.

  • No previous occurrences were identified as applicable to the circumstances of this event.

G. COMPONENT FAILURE DATA

Failed Equipment: N/A Component Manufacturer: N/A Component Model Number: N/A Component Part Number: N/A This event has not been reported to ICES since there was no equipment failure.

05000265/LER-2016-00224 June 2016Quad Cities

Motor Operated (MO) HPCI Outboard Main Steam Isolation Valve (MO 2-2301-5). The packing leak was causing a two (2) foot steam plume to impinge on the valve limit switch compartment, potentially impacting the motor operator for the MO 2-2301-5 valve.

Due to the uncertainty on how the steam impingement would affect the valve limit switch compartment, Operations conservatively isolated the steam leak by closing the HPCI Inboard Main Steam Isolation Valve (MO 2-2301-4). With the steam supply isolated, HPCI was declared inoperable and Technical Specification (TS) 3.5.1 Condition G was entered.

The cause of the packing leak was a non-modern style packing installed in 2007 to repack valve MO 2-2301-5.

This packing material was susceptible to premature degradation.

Corrective actions included repacking the valve with modern packing and performance of valve diagnostic testing.

The safety significance of this event was minimal. Given the impact on the HPCI system, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(D), which requires the reporting of 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.

APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

05000265/LER-2016-00119 May 2016Quad Cities

On March 21, 2016, at 2200 hours, with Unit 2 shutdown for refuel outage Q2R23, the as-found local leak rate tests (LLRT) for the four (4) main steam lines (MSL) were performed following closure of the main steam isolation valves (MSIV). The initial as-found LLRT on the "A" and "C" MSL MSIVs exceeded the minimum pathway criteria (lesser leakage in a line) of the Technical Specifications (TS), and the combined total leakage of all MSLs also exceeded the minimum pathway criteria (lesser leakage in each line when combined for all MSIVs) of the TS.

Corrective actions included flushing, disassembling, inspecting, repairing, and retesting the valves. Future corrective actions include installation of an improved spherical nose plug design to the MSIV plug and seat, and installation of an anti-rotation device to the MSIV pilot.

Valves 2-0203-2A and 2-0203-1C were disassembled and inspected. The most likely cause for the higher than expected leakages has been determined to be a valve design that is susceptible to a degraded main plug / seat interface during valve closure. A contributing cause was susceptible pilot plug / seat misalignment, due to pilot disc stem nut wear.

The safety significance of this event was minimal. The total primary containment leakage of 315.866 scfh was well within the allowed leakage limit of 1372.99 scfh (La). However, since the "A", "B", "C" and "D" MSL MSIV as-found leakage exceeded the TS limit, and the combined total leakage of all MSLs exceeded the TS limit, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of a past operation or condition which was prohibited by the plant Technical Specifications.

05000254/LER-2016-00215 March 2016Quad Cities

building. The alarms occurred during an entry to the Unit 2 Reactor Water Cleanup (RWCU) pump room. A negative reactor building pressure was restored within two minutes (approximately 20:40 hours) without operator action.

Since both Units 1 and 2 share a common reactor building (RB), the loss of differential pressure impacted both Units 1 and 2 secondary containments.

The cause was a sheared air line inside the Unit 1 RB ventilation exhaust plenum which depressurized the air header supplying operating air to all three Unit 1 reactor building exhaust fan isolation dampers, causing the dampers to fail open, including the one on the standby fan.

Corrective actions included replacing the sheared air line, and the addition of a preventive maintenance task for replacement of equivalent air lines on all RB supply and exhaust fan dampers.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10CFR 50.73(a)(2)(v)(C), which requires the reporting of 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.

05000254/LER-2016-00111 March 2016Quad Cities

building. The alarms occurred during an entry to the Unit 2 Reactor Water Cleanup (RWCU) Heat Exchanger (HX) room. A negative reactor building pressure was restored within one minute (alarm cleared at 13:41 hours) by immediately securing a reactor building supply fan. Since both Units 1 and 2 share a common reactor building (RB), the loss of differential pressure (RB pressure went positive) for approximately one (1) minute impacted both Units 1 and 2 secondary containments.

The cause was a sheared air line inside the Unit 1 RB ventilation exhaust plenum which depressurized the air header supplying operating air to all three Unit 1 reactor building exhaust fan isolation dampers and causing the dampers to fail open, including the one on the standby fan.

Corrective actions included replacing the sheared air line, and the addition of a preventive maintenance task for replacement of equivalent air lines on all RB supply and exhaust fan dampers.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10CFR 50.73(a)(2)(v)(C), which requires the reporting of 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.

05000254/LER-2015-0105 February 2016Quad Cities

system train "B" in support of planned maintenance on the non-safety related train "A". The CREV train "B" did not start. Once the CREV train "B" system failed to start, the non-safety related train "A" was restarted. Proper operation of the CREV system and CREV air conditioning (NC) systems could not be ensured, so both CREV and CREV AC system were declared inoperable. As a result, Technical Specification (TS) 3.7.4, Condition A, and TS 3.7.5, Condition A were entered.

The cause of the CREV train "B" system failure to start was the differential pressure switch, which is installed in a vibration susceptible location on the ductwork. This differential pressure switch makes up the interlock between the CREV train "B" system and the non-safety related Control Room HVAC train "A". The differential pressure switch's normally open contacts had temporarily welded together not allowing the starting signal to be received by the CREV train "B" system.

Corrective actions included replacing the differential pressure switch. A future corrective action will relocate the differential pressure switch off of the ductwork to minimize the vibration effects.

The safety significance of this event was minimal. Given the impact on the CREV system, which is common to both units, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(D), which requires the reporting of 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.

05000254/LER-2015-00924 September 2015Quad Cities

Emergency Ventilation (CREV) System, technicians noticed that the 'B' Air Filtration Unit (AFU) Booster Fan discharge damper became stuck in a partially open position following the 'B' AFU Booster Fan trip on a high ammonia input signal. Due to the uncertainty of being able to achieve rated airflow for the CREV System caused by recirculation that would result from running the 'A' AFU Booster Fan with the 'B' Booster Fan discharge damper partially stuck open, the CREV System was declared inoperable and surveillance testing was secured. As a result, Technical Specification 3.7.4, Condition A, was entered.

The cause of the damper failure was due to inadequate clearance between the damper seat sealing area and the damper blade to shaft fasteners which resulted in misalignment which caused the damper shaft to bend and the damper to become stuck.

Corrective actions included securing the stuck damper in the closed position and evaluating all CREV System dampers. A preventative maintenance task will be developed to stroke and inspect the dampers to identify binding; modifications will be made if necessary to other CREV System dampers to prevent binding.

The safety significance of this event was minimal. Given the impact on the CREV System, which is common to both Units, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(D), which requires the reporting of 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.

05000254/LER-2015-00817 August 2015Quad Cities

On June 19, 2015, at 0153 hours, the control room was notified that the Unit 1/2 emergency diesel generator (EDG) room manual side door and the Unit 1 reactor building (RB) side door, as a part of the personnel interlock to the 1/2 EDG room, were opened simultaneously. The failure of this interlock caused a temporary loss of secondary containment (inoperable) per Technical Specification (TS) 3.6.4.1, Condition A. The 1/2 EDG manual interlock door was closed immediately, and the secondary containment boundary was immediately reestablished. Operators verified the RB (secondary containment) differential pressure remained negative during this event.

Secondary containment remained available and functional during the event since the secondary containment interlock was immediately restored by closing the 1/2 EDG room manual side door, and since the RB differential pressure was maintained during the event. The RB is a common volume to both Units 1 and 2, and an interlock failure can impact the secondary containment for both units.

The event involved the mechanical latch on the 1/2 EDG room door that had failed to fully engage into the strike, although it did engage far enough to make up the electrical logic. The apparent cause of the interlock failure was determined to be the 1/2 EDG room door failed to remain closed due to inadequate design of the interlocks.

Corrective actions included troubleshooting and administratively controlling the interlock door. The preventive maintenance frequency of the interlock door will be increased, the door will be replaced, and a modification will be performed to address single point vulnerabilities.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of 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.

05000254/LER-2015-00724 July 2015Quad Cities

alarm. A fire damper inspection was being performed that opened a Control Room HVAC ductwork access hatch that caused the alarm. The hatch was opened and immediately shut, re-establishing the boundary of the Control Room Envelope (CRE). The Control Room Emergency Ventilation (CREV) system was declared inoperable due to opening the ventilation duct hatch without prior administrative controls in place. As a result, Technical Specification 3.7.4, Condition C, was entered and subsequently exited within approximately one minute.

The cause of the inadvertent CRE breach was the design drawing contained in the work package that was reviewed during the Plant Barrier Impairment (PBI) screening did not adequately define the boundaries of the CRE.

Corrective actions included reviewing all open PBI packages. Associated Control Room boundary drawings and procedures will be revised to correctly annotate the proper CRE boundary to include the MCR ventilation ductwork access hatch.

The safety significance of this event was minimal. Given the impact on the MCR envelope, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(D), which requires the reporting of 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.

05000254/LER-2015-0051 June 2015Quad Cities

On April 2, 2015, at 1810 hours, with Unit 1 in Mode 1 and operating at 100% power, an alarm was received in the control room indicating that two of three turbine throttle pressure transmitters on Unit 1 had failed low.

Reactor power was reduced to 20% and the main turbine was secured.

A steam leak was confirmed to have originated between the turbine throttle pressure transmitter sensing line isolation valve and the 30 inch main steam D-Ring header. Since the steam leak could not be isolated with Unit 1 at power, a manual scram was inserted on Unit 1 at 2133 hours, and the main steam isolation valves (MSIVs) were manually closed. A forced outage was initiated to investigate and perform repairs.

The reactor and turbine responded as designed.

Operators performed required actions safely and in accordance with procedures and training.

Operators needed to use relief valves to control reactor pressure since the turbine bypass valves were unavailable for normal pressure control after closing the MSIVs to isolate the steam leak at the D-ring. A drywell pressure rise occurred due to a relief valve tailpipe vacuum breaker which had inadvertently stuck open; the relief valve was then closed, a drywell cooler was started, and drywell pressure returned to normal.

The root cause of the failure of the sensing line was determined to be inadequate monitoring of the sensing line supports that allowed each of the clamps to degrade and loosen over time resulting in high cycle fatigue cracking and eventual fracture of the sensing line.

Corrective actions included replacing the 3/4 inch carbon steel section of the failed sensing line, including the isolation valve, and replacing the associated degraded supports. Future corrective actions include conducting focused inspections of small bore piping/tubing on both accessible and inaccessible systems.

The safety significance of this event was minimal. This report is submitted in accordance with 10 CFR 50.73 (a)(2)(iv)(A) which requires the reporting of any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B).

05000254/LER-2015-00420 May 2015Quad Cities

On 03/21/15, at 0800, with Unit 1 shutdown for refuel outage Q1R23, the Unit 1 Off-Line Automatic Blowdown Logic Test was being performed when two Electromatic Relief Valves (ERVs) failed to actuate since the 'N trip logic for the automatic function of the Automatic Depressurization System (ADS) did not energize.

Maintenance burnished the 'A' trip logic relay contacts and the 'B' trip logic relay contacts. The as-found data on the 'B' trip logic of ADS was lost during this troubleshooting, and it is unknown if the 'B' trip logic of the automatic function would have functioned. Once the relay contacts were burnished, the continuity was checked and found acceptable.

The apparent cause for the loss of the 'A' trip logic system was due to relay contact oxidation buildup. The oxidation buildup was due to inadequate preventative maintenance instructions.

The immediate corrective action consisted of burnishing the relay contacts. Future corrective actions include updating the preventative maintenance procedure and replacing the two relays on the 'A' trip logic system that had the oxidation buildup.

This report is submitted in accordance with 10 CFR 50.73 (a)(2)(v)(D) which requires the reporting of 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; and 10 CFR 50.73 (a)(2)(vii)(D) which requires the reporting of any event where a single cause or condition caused two independent channels to become inoperable in a single system designed to mitigate the consequences of an accident.

05000254/LER-2015-00329 April 2015Quad Cities

On March 2, 2015, at 2359 hours, with Unit 1 shutdown for refuel outage Q1 R23, the as-found local leak rate tests (LLRT) for the four (4) main steam lines (MSL) were performed following closure of the main steam isolation valves (MSIV). The initial as-found LLRT on the "D" MSL MSIVs exceeded the minimum pathway criteria (smaller leakage in a line) of the Technical Specifications (TS), and the combined total leakage of all MSLs also exceeded the minimum pathway criteria (smaller leakage in each line when combined for all MSIVs) of the TS.

The leaking valves were disassembled and inspected. Although some seat ring wear is a normal occurrence in these valves, the excessive leakage was associated with wear due to the high friction that develops between the plug and the seat as initial contact occurs on valve closure, followed by the sliding action as the plug centers itself in the seat, which is inherent with the "Y"-style globe valve.

The most likely cause for the higher than expected leakages has been determined to be a valve design that is susceptible to localized seat wear during valve closure. The plug tends to drag across the sharp edge of the seat ring, and over time, if the closure strokes allow the plug to drag along the same contact point then wear will occur on the sharp edge of the seat ring.

Corrective actions included flushing, disassembling, inspecting, repairing, and retesting the valves. Future corrective actions include installation of an improved spherical nose plug design to the MSIV plug and seat.

The safety significance of this event was minimal. The total primary containment leakage of 283.59 scfh was well within the allowed leakage limit of 823.79 scfh (0.6La). However, since the "D" MSL MSIV as-found leakage exceeded the TS limit, and the combined total leakage of all main steam lines exceeded the TS limit, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of a past operation or condition which was prohibited by the plant Technical Specifications.

05000254/LER-2015-0026 April 2015Quad Cities

On February 10, 2015, at 1055 hours, a maintenance individual notified the control room that both doors in the secondary containment interlock between the Reactor Building (RB) Unit 1 High Pressure Coolant Injection (HPCI) room and the Unit 1 Turbine Building (TB) were opened simultaneously. The failure of this interlock caused a loss of secondary containment (inoperable) per Technical Specification (TS) 3.6.4.1, Condition A. The RB-side door was immediately reclosed, and the secondary containment boundary was reestablished. Operators verified the RB (secondary containment) differential pressure remained negative during this event.

Secondary containment remained available and functional during the event since the secondary containment interlock was immediately restored by closing the RB-side door and the RB differential pressure was maintained during the event. No RB low differential pressure alarms were received during this event. The RB is a common volume to both Units 1 and 2, and an interlock failure can impact the secondary containment for both units.

The cause of the interlock failure was due to a bent locking bolt with insufficient strength to withstand the standard practice of challenging that fire doors are locked closed after passing through them. The bent locking bolt caused its door plungers to not engage allowing its door to open while the opposite door in the interlock was being opened.

Corrective actions included repairing the bent locking bolt on the TB-side passive door and realigning the TB-side doors to ensure the plungers would be functional. The locking bolt will be replaced with a higher strength assembly.

The safety significance of this event was minimal.

Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of 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.

05000265/LER-2015-00130 March 2015Quad Cities

On March 5, 2015, at approximately 1640, the watertight door for the Unit 1 High Pressure Coolant Injection (HPCI) room was found open with no person in attendance. In this condition, the door is not able to perform the flood protection function. With no person in attendance the door would not be shut to prevent internal flood water from entering the Unit 1 HPCI room. This condition would result in the inoperability of equipment in the room it is designed to protect from flooding.

The construction of the adjacent Unit 1 and Unit 2 HPCI rooms provides no flood barrier between the two rooms.

Therefore, a condition that results in flood protection being nonfunctional to one HPCI room also has an effect on the opposite Unit HPCI. The Unit 1 HPCI watertight door being found open, with no one in attendance, results in the unplanned inoperability of the Unit 2 HPCI, since the Unit 2 HPCI is required to be operable by Technical Specifications in Mode 1. The Unit 1 HPCI was not required to be operable since Unit 1 was in Mode 5. Therefore, this Licensee Event Report is being submitted in accordance with 10 CFR 50.73 (a)(2)(v)(D) for an event or condition that could have prevented fulfillment of a safety function.

05000254/LER-2015-0016 March 2015Quad Cities

On January 6, 2015, Electrical Maintenance was preparing for planned maintenance activities at a 480V Motor Control Center (MCC). One of the technicians identified that the breaker in cubicle Al was in the tripped position. Breaker Al is the Unit 2 power supply breaker to the Unit 0 Fuel Oil Transfer Pump (FOTP) for the Unit 0 Emergency Diesel Generator (EDG).

Troubleshooting identified the cause of the breaker trip was due to high resistance contacts on the HGA power transfer relay. This relay was replaced and tested satisfactory on January 8, 2015.

This breaker most likely tripped under load, which would have occurred during the Unit 0 EDG 24 hour endurance run when the EDG was loaded to Unit 2 on December 30, 2014. Since planned maintenance occurred on the Unit 0 EDG prior to the endurance run, the time of inoperability of the Unit 0 EDG started on December 29, 2014 and ended when the failed relay was replaced and tested satisfactory on January 8, 2015, for a total of 10 days. This exceeded the allowed outage time of Technical Specifications 3.8.1 for one EDG inoperable. Therefore, this Licensee Event Report is being submitted in accordance with 10 CFR 50.73 (a)(2)(i)(B) for a condition prohibited by Technical Specifications.

NRC FORM 368 (02-2014)

05000254/LER-2014-00513 February 2015Quad Cities

On December 15, 2014 at 0730 hours, the south Main Control Room (MCR) door was unable to be fully closed due to a failure of the closer mechanism. Technical Specification (TS) 3.7.4, Condition C, was entered due to the inoperable Control Room Envelope (CRE). A security guard was stationed to provide controlled access to the MCR.

The CRE function was restored shortly after the event when the door closer arm linkage was disconnected, the door was confirmed closed, and successfully smoke tested.

The cause of the MCR door closer mechanism failure was a manufacturing defect of the pinion gear.

Corrective actions included replacement of the door closer mechanism, and a preventative maintenance task to replace the closer mechanism will be established.

The safety significance of this event was minimal. Given the impact on the MCR envelope, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(D), which requires the reporting of 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.

05000254/LER-2014-00416 January 2015Quad Cities

turbine building (TB) doors for the 647 foot elevation secondary containment interlock were opened simultaneously. The failure of this interlock caused a loss of secondary containment per Technical Specification (TS) 3.6.4.1, Condition A. The doors were immediately closed and secondary containment pressure remained negative throughout the event. The Field Supervisor was dispatched to investigate the interlock issue and found that the issue was intermittent. Operators staged signage and ropes to administratively control the interlock closed until repairs were performed.

Secondary containment remained available during the event because the secondary containment interlock was immediately restored by closing the doors, and the RB differential pressure was maintained throughout the event. No RB low differential pressure alarms were received during the event. The RB is a common volume to both Units 1 and 2, and an interlock failure can impact the secondary containment for both units.

The cause of the interlock failure was the door magnets on the Unit 2 TB door did not fully engage to hold the door closed due to inadequate design of the interlocks.

Corrective actions included adjustment of the lower magnet on the TB door to ensure good connection between the door and the frame. A modification to the 647 foot elevation interlock doors that addresses single point vulnerabilities will be implemented, and the preventative maintenance frequency will be shortened from 12 months to 6 months.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of 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.

05000254/LER-2014-00317 July 2014Quad Cities

On May 22, 2014, at 2150 hours, a mechanic notified the Main Control Room that both doors (DR) in the secondary containment (NG) interlock (IEL) to the High Pressure Coolant Injection (HPCI) room from the Turbine Building (TB) (NM) were opened simultaneously. The failure of this interlock caused a loss of secondary containment per Technical Specification (TS) 3.6.4.1, Condition A. The doors were immediately reclosed, and the secondary containment boundary was reestablished. Operators verified the Reactor Building (RB) (secondary containment) differential pressure was maintained operable at greater than 0.10 inch of vacuum water gauge.

Secondary containment remained available and functional during the event since the secondary containment interlock was immediately restored by closing the doors and since the RB differential pressure was maintained during the event. No RB low differential pressure alarms were received during this event. The RB is a common volume to both Units 1 and 2. An interlock failure can impact the secondary containment for both units.

The cause of the interlock failure was due to a bent locking bolt resulting in misalignment of the interlock plungers on the TB-side door. The mechanical interlock device could be defeated inadvertently in this condition.

Corrective actions included replacing the bent locking bolt and realigning the TB-side doors.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of 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.

05000265/LER-2014-00315 July 2014Quad Cities

On May 16, 2014, with Unit 2 at full power, a main condenser flow reversal was performed. During the flow reversal, anomalous indications were noted for the 2C condenser (COND) backpressure response. As a result of the slow response, RPS pressure switch (PS) 2-0503-B was declared inoperable. This pressure switch is one of the inputs into the Reactor Protection System (RPS) (JD) for the Turbine Condenser Vacuum-Low function. The isolation valve (RTV) for this pressure switch was found partially closed. The valve was opened and normal indication was restored.

During the review of this event, it was identified that the RPS pressure switch was inoperable since May 6, 2014. Unit 2 did not meet the required number of channels per TS 3.3.1.1, Function 10, Turbine Condenser Vacuum-Low, when transitioning to MODE 1 at 2252 on May 6, 2014, as required by TS 3.0.4, and did not take the actions required in the allowed outage time.

The safety significance of this event was minimal. Sufficient redundant condenser backpressure instrumentation was operable to maintain scram capability and the RPS safety function. This event is reportable as an operation or condition prohibited by plant Technical Specifications per 10 CFR 50.73(a)(2)(i)(B).

05000265/LER-2014-0022 June 2014Quad Cities

On April 2, 2014, at 1228 hours, a Fire Alarm System (FAS) alarm was received for the Unit 2 D heater bay area. Although entry into the room at the time identified only a steam leak, subsequently various spurious alarms and electrical system anomalies occurred.

At 1303 hours, Unit 2 was manually scrammed, the turbine was tripped, and the main steam isolation valves (MSIVs) were closed to ensure the steam leak was isolated. A fire was identified to have occurred in the D heater bay (an area of the plant containing the high pressure (final stage) D feedwater heaters, and several Unit 2 cable trays and risers). The fire was extinguished by the automatic wet pipe sprinkler fire suppression system.

At 1340 hours, due to the manual de-energizing of safety-related motor control center (MCC) 29-1 in the reactor building in response to notification that smoke had been observed, an ALERT level Emergency Action Level classification was declared as HA3 (fire in a vital area affecting safety system equipment). The emergency was terminated at 2132 hours.

The cause of the event was an existing cable flaw that was caused by cable routing that exceeded the required minimum static bend radius.

Corrective actions included repairing impacted cables, replacing the failed steam seal expansion joint, operating procedure revisions, and additional inspections/tests.

The safety significance of this event was minimal. Given the impact on multiple systems, this report is submitted in accordance with 10 CFR 50.73 (a)(2)(iv)(A) for manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B); in accordance with 10 CFR 50.73 (a)(2)(v)(D) for an event that could have prevented the fulfillment of the safety function of systems needed to mitigate the consequences of an accident; and in accordance with 10 CFR 50.73(a)(2)(i)(A), for the completion of a nuclear plant shutdown required by the plant's Technical Specifications.

05000254/LER-2014-0022 June 2014Quad Cities

On April 1, 2014, at 1357 hours, an Instrument Maintenance Technician notified the Main Control Room that both doors in the secondary containment interlock on the 595 foot elevation between the Reactor Building (RB) and the Unit 2 Reactor Feed Pump room (located inside the turbine building (TB)) were opened simultaneously. The failure of this interlock caused a loss of secondary containment per Technical Specification (TS) 3.6.4.1, Condition A. The doors were immediately reclosed, and the secondary containment boundary was immediately reestablished.

Operators verified the RB (secondary containment) differential pressure was maintained operable at greater than 0.10 inch of vacuum water gauge.

Secondary containment remained available and functional during the event since the secondary containment interlock was immediately restored by closing the doors and since the RB differential pressure was maintained during the event. No RB low differential pressure alarms were received during this event. The reactor building is a common volume to both Units 1 and 2, and an interlock failure can impact the secondary containment for both units.

The cause of the interlock failure was due to a malfunctioning interlock door hydraulic actuator and time delay relays had allowed the second door to open before the first door was secured.

Corrective actions included replacing the failed actuator and adjusting the limit switch. A set point change will be implemented to resolve relay time delay issues.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of 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.

05000265/LER-2014-00130 May 2014Quad Cities

On March 31, 2014, at 1302 hours, an Inservice Inspection Program VT-2 examination of the Unit 2 Control Rod Drive (CRD) Hydraulic Control Unit (HCU) ASME Class 2 piping and components was being performed. An apparent through-wall valve body leak of approximately two drops per minute was discovered on the 2-0305-101-18-27 CRD HCU Scram Insert Isolation Valve. This valve is subjected to full reactor pressure during normal service and during this inspection. This valve is the isolation valve to the reactor vessel CRD drive housing, and since it is the first isolation boundary off of the reactor vessel, it therefore cannot be isolated from the reactor coolant system to allow repairs. The valve was declared inoperable, Technical Specifications LCO 3.4.4 Condition C was entered, and the Unit was shutdown and depressurized to effect repairs.

On April 1, 2014, the 2-0305-101-18-27 valve was removed from the system and shipped for analysis. It was determined that the through wall leak that developed was the direct result of an inherent manufacturing defect that eventually propagated to the valve surface following years of pressure and temperature cycles that the system normally experiences.

Corrective actions included replacing the failed isolation valve and performing additional CRD system inspections. A root cause analysis was performed and no additional contributing factors were identified.

The safety significance of this event was minimal since the leakage rate was very small and full scram capability was maintained by the control rod. Due to the impact on the reactor coolant pressure boundary, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(ii)(A), which requires the reporting of any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. Since a plant shutdown was completed as required by the plant Technical Specifications, this report is also submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(A), which requires the reporting of the completion of any nuclear plant shutdown required by the plant's Technical Specifications.

05000254/LER-2014-0015 May 2014Quad Cities

On March 4, 2014, at 1917 hours, a Fuel Pool Channel 1B High Radiation alarm occurred. The 1B fuel pool radiation monitor spiked high at 50 mRem/hr, trended downward, and returned to its previous steady-state value of 14 mRem/hr. The high spike caused the Unit 1 and Unit 2 Reactor Building (RB) ventilation and Control Room (CR) ventilation to isolate as designed. The Standby Gas Treatment System (SBGTS) was already in operation for a scheduled surveillance as of 1900 hours on March 4, 2014.

The Unit 1 and Unit 2 RB ventilation system isolation caused the secondary containment differential pressure to be momentarily lost (pressure went positive), and secondary containment was declared inoperable. The RB ventilation system was then shut down for scheduled maintenance, the CR ventilation system was returned to its normal configuration, and troubleshooting was initiated.

Since both Units 1 and 2 share a common RB, the loss of differential pressure (RB pressure went positive) for approximately three (3) minutes impacted both Units 1 and 2 secondary containments.

The cause of the secondary containment loss of differential pressure event was due to a failed fuel pool radiation monitor detector that caused the RB ventilation system to isolate. The detector failure was caused by a manufacturing defect that caused double pulsing on the GM-Tube.

Corrective actions included replacing the failed radiation monitor detector, and sending the failed detector to the vendor for failure analysis.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of 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.

05000254/LER-2013-00328 May 2013Quad Cities

On March 26, 2013, at 1541 hrs, while Unit 1 was shutdown for refueling outage Q1R22, leakage was identified exiting from the 2-inch reactor vessel head vent piping during a Reactor Pressure Vessel (RPV) pressure test. The leakage rate was approximately 20 drops per minute (dpm). The RPV head vent line provides a connection from the RPV upper steam dome volume to provide a vent path to the drywell sumps during shutdown conditions and to the main steam lines to vent non- condensable gasses during power operation.

The leakage originated from a socket weld (designated as FW14) between the pipe and a 90 degree fitting. The RPV pressure test was stopped and the reactor vessel depressurized to allow additional inspections and necessary repairs.

The most probable cause of the leakage was determined to be a defect in the socket weld attributed to porosity and/or slag, originating from initial construction in 1970, prior to the initial start-up of Unit 1.

Corrective actions included repairing the weld by removal of the existing weld material and application of a new weld. Future corrective actions include inspecting other similar RPV head vent welds and repairing/evaluating as necessary.

The safety significance of this event was minimal since the leakage rate was very small and was found while the reactor was shutdown. Given the impact on the reactor coolant pressure boundary, this report is submitted in accordance with the requirements of 10 CFR 50.73 (a)(2)(ii)(A), which requires the reporting of any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

05000254/LER-2013-00210 May 2013Quad Cities

On March 11, 2013, during refueling outage Q1R22 at approximately 0750 hours, the Unit 1 Main Steam Isolation Valve (MSIV) as-found stroke time test results indicated that all four outboard MSIVs failed to close within the required Technical Specification (TS) time of less than or equal to five (5) seconds. The four inboard MSIVs, tested just prior to the outboard MSIVs, all closed within the required TS time of less than or equal to five (5) seconds. The safety significance of this event was minimal.

The cause of the Unit 1 outboard MSIV slow closure timing is the actuator seals had degraded due to age and wear.

Corrective actions included replacing the 1-0203-2B MSIV actuator, readjusting as-left closure times on all outboard MSIVs, and satisfactory retesting prior to startup. Follow-up corrective actions include replacing the 1-0203-2A, 2C, and 2D MSIV actuators and increasing the preventive maintenance frequency.

Since all four outboard MSIVs failure to stroke within the required times was attributed to degraded actuator seals, it is likely the degradation occurred over time since the last successful refueling outage testing and during power operations when the required TS 3.6.1.3, Primary Containment Isolation Valves (PCIV5), was applicable.

Therefore this condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), which requires reporting of any operation or condition that was prohibited by the plant's Technical Specifications.

05000254/LER-2013-00130 April 2013Quad Cities

On March 1, 2013, during restoration of the 1B Core Spray (CS) logic test, the lineup for the 1/2 Emergency Diesel Cooling Water Pump (1/2 EDGCWP) was to be restored to normal. It was discovered that the 1/2 EDGCWP was lined up to Unit 2 instead of the required Unit 1. At the time of discovery all required systems were operable. However, prior to discovery, the Unit 1 Emergency Core Cooling System (ECCS) room coolers were inoperable since Technical Specifications (TS) 3.7.2, Diesel Generator Cooling Water (DGCW) System, Action B.1, was not met because an operable EDGCW subsystem was not "aligned" to the ECCS room coolers, and associated ECCS were not alternatively declared inoperable per Action B.2.

During the time the 1/2 EDGCWP isolation valve to Unit 1 ECCS room cooling (1/2-3999-89) was improperly aligned, it was however, available to be opened via Operator manual actions had an ECCS initiation occurred. The safety significance of this event was minimal since the Unit 1 EDGCWP and the 1/2 EDGCWP were available and capable of supporting Unit 1 ECCS Room cooling.

The cause of the event was the individual worker's lack of execution of proper human performance tools.

Corrective actions included the addition of immediate oversight of operations briefs and direct documented supervision of equipment operators during the upcoming refueling outage. Planned corrective actions include implementation of the process of focused intervention directed at the crew level based on Employee Observation System trending and analysis data, and documenting the performance of the human performance "Out-of-the-Box" (OBE) Evaluations pertaining to use of human performance tools.

Given the impact of the misaligned 1/2 EDGCWP on the operability of Unit 1 ECCS Room cooling for CS, Residual Heat Removal (RHR), and High Pressure Coolant Injection (HPCI) systems, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of any operation or condition which was prohibited by the plant's Technical S ecifications. -

05000265/LER-2012-00314 June 2012Quad Cities

On April 18, 2012, at 1511 hrs, while at low power conditions after refueling outage Q2R21, an automatic reactor scram occurred on Unit 2 due to high reactor pressure. The pressure increase occurred during post-modification testing on the main generator automatic voltage regulator (AVR) which had been upgraded during refueling outage Q2R21. The testing included a generator load reject, which was in progress when the pressure transient occurred. There were no complications during the reactor scram and subsequent turbine trip, and all systems functioned as required. Operators performed required actions safely and in accordance with procedures and training.

The cause of the automatic scram was due to high reactor pressure created by the load rejection associated with the main generator voltage regulator testing, coincident with unresponsive opening demand of turbine control valves (TCVs) that impacted the turbine bypass valves (TBVs) ability to control reactor pressure. Since the digital electro hydraulic control (DEHC) system design lacked the required Intercept (IV) EHC shutoff valves, this resulted in low EHC pressure and caused the TCVs to be unresponsive. Unit 1 was unaffected by the event and remained at 100% power.

Corrective actions included evaluating the impact of the event on the operability of the TBVs, and applying a Minimum Critical Power Ratio (MCPR) Operating Penalty (TS 3.7.7) when reactor power is between 25% and 50%. Future corrective actions include development of a hardware/software modification to the DEHC system to correct the design deficiency.

The safety significance of this event was minimal. This event is reportable (Unit 2) per 10 CFR 50.73(a)(2)(iv)(A), which requires the reporting of any event or condition that resulted in manual or automatic actuation of the reactor protection system (RPS), including reactor scram; (Units 1 and 2) 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of any operation or condition which was prohibited by the plant's Technical Specifications; and (Units 1 and 2) 10 CFR 50.73 (a)(2)(v)(C), which requires the reporting of any event or condition that could have prevented the fulfillment of the safety function of structures or systems needed to control the release of radioactive material.

05000265/LER-2012-0024 June 2012Quad Cities

On April 4, 2012, at 1716 hrs, while Unit 2 was shutdown for refueling outage Q2R21, leakage was identified exiting from a 2­ inch reactor vessel instrumentation nozzle (N-11B) during a Reactor Pressure Vessel (RPV) pressure test. The leakage amount was approximately 60 drops per minute (dpm). The vessel penetration (N-11B) provides the connection point for the reference leg of the "B"-train of the Reactor Vessel Level Instrumentation System (RVLIS).

The leakage originated from the area where the nozzle penetrates the vessel wall. The nozzle is welded on the inside of the vessel, so the actual attachment weld could not be examined.

The RPV pressure test was stopped and the reactor vessel depressurized to allow additional inspections and necessary repairs.

The most probable cause of the leakage was determined to be lntergranular Stress Corrosion Cracking (IGSCC) that was likely influenced by higher residual stresses that remained in the nozzle assembly following nozzle replacement in 1970, prior to the initial start-up of Unit 2.

Corrective actions included repairing the nozzle with IGSCC resistant material, and obtaining approval of a Relief Request from the NRC prior to startup to allow the flaw to remain for one operating cycle. Future corrective actions include inspecting other similar RPV nozzles, and performing a specialized flaw evaluation to support safe operation for continued operating cycles.

The safety significance of this event was minimal given the leakage was very small, was found while the reactor was shutdown, and if leaked during plant operation, did not exceed Technical Specification (TS) leakage limits for unidentified drywell leakage.

Given the impact on the reactor vessel pressure boundary, this report is submitted in accordance with the requirements of 10 CFR 50.73 (a)(2)(ii)(A), which requires the reporting of any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

05000254/LER-2011-0036 September 2011Quad Cities

On July 6, 2011 at 1410 hours, the Unit 1/2 (common) "B" Control Room Emergency Ventilation (CREV) Refrigeration Condensing Unit (RCU) was declared Inoperable due to an increase in vibration amplitude caused by a broken chiller compressor connecting rod. Operators performed required actions to safely secure the CREV RCU in accordance with procedures and training, and without complications. This resulted in entering Technical Specification 3.7.5, Condition A (30 day Action). This event affected both the Unit 1 and Unit 2 Control Rooms since they share a common control room and CREV system.

The Train B CREV Air Conditioning (AC) system is a single train safety-related system that is designed to operate in a post- accident condition to maintain design temperature in the Control Room Envelope (CRE), and loss of the CREV AC could impact the plant's ability to mitigate the consequences of an accident.

The CREV RCU chiller compressor connecting rod failure was caused by flooded starts of the compressor. The root cause of the flooded start was an unanticipated failure mode for the design application of the system.

Corrective actions included replacing the failed compressor with a new compressor, and revising the frequency for vibration analysis on the B CREV RCU to monthly. Future corrective actions include installation of an automatic pump-down modification for the B CREV RCU compressor.

The safety significance of this event was minimal.

Given the impact on the CREV AC system, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of 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.

05000254/LER-2011-00212 August 2011Quad Cities

On June 13, 2011, Quad Cities Unit 1 was performing startup from refueling outage Q1R21. During power ascension while at 61% power (560 MWe) a steam leak was identified on the main steam line downstream of turbine control valve #1 (1-5699-CV1'). Reactor power was reduced and the reactor was manually scrammed. The steam leak occurred at a recently repaired sensing line that had previously leaked. A forced outage was initiated to investigate and performs repairs.

The reactor and turbine responded as designed. Operators performed required actions safely and in accordance with procedures and training. There were no complications during the reactor scram and turbine trip, and all systems functioned as required.

The root cause for the manual scram due to a steam leak from a pressure sensing line repair was that the capped pipe stub repair failed due to a fatigue induced flaw that was not known to exist at the time of the repair.

Corrective actions included replacing the Unit 1 sensing line sock-o-lets with gamma plugs. Future corrective actions include replacement of the equivalent Unit 2 sensing line sock-o-lets with gamma plugs.

The safety significance of this event was minimal. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A), as any event or condition that resulted in manual or automatic actuation of the reactor protection system (RPS), including reactor scram.

05000254/LER-2011-00120 June 2011Quad Cities

On April 19, 2011, at 1430 hours, a water leak was discovered on the Unit 1 Emergency Diesel Generator Cooling Water Pump (EDGCWP) room cooler. As a result, the Unit 1 EDG (Division II) was declared inoperable and Technical Specification (TS) 3.8.1 was entered. (ENS Report No. 46769) At this time, Division I Residual Heat Removal (RHR) was inoperable for planned maintenance on the Division I RHR room cooler. In the event of a LOOP/LOCA scenario, RHR would not have been able to perform its safety function (i.e., Division I RHR was impacted by the ongoing room cooler maintenance and Division II RHR would not have had an operable on-site emergency power source). Note that at the time of discovery, offsite power was available to Division II RHR as well as the Division I and Division II Core Spray systems.

At 1816 hours, the Division I RHR system room cooler was restored and Division I RHR was declared operable.

The room cooler leak was caused by under-deposit corrosion on the piping due to raw water conditions. The through wall leak on the cooler was subsequently repaired and the Unit 1 EDGCWP room cooler was returned to operable status on April 21, 2011.

Given the impact on the RHR system, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of 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.

05000265/LER-2011-00110 March 2011Quad Cities

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On January 12, 2011, at 1020 hours, the Unit 2 Essential Service (ESS) 480V Bus 29 (BU) was inadvertently de-energized. The cause of this bus trip was due to inadvertent contact with the bus feed breaker local trip pushbutton (JS) by a station employee during unrelated work activities in the bus feed breaker area. While Bus 29 was de-energized, Division II core and containment cooling systems (BO) were unavailable and inoperable; however, normal power to Bus 29 was restored within 6 minutes.

The plant responded as designed to the loss of Bus 29, with the exception that the normally supplied reactor building ESS 480V Bus Motor Control Center (MCC) 28/29-5 did not receive the auto transfer of supplied power from its reserve feed 480V Bus 28.

Since either 480V Bus 29 or Bus 28 can feed Bus 28/29-5, this condition resulted in a loss of power to Bus 28/29-5 and rendered both divisions of the Low Pressure Coolant Injection (LPCI) (BO) mode of the Residual Heat Removal (RHR) (BO) system inoperable. Therefore, Technical Specification 3.5.1.E was entered, requiring restoration of LPCI within 72 hours.

It was subsequently determined that the "M" auxiliary contactor (CNTR) from Bus 29 had failed which caused the auto transfer logic from Bus 29 to Bus 28 to fail.

Bus 28/29-5 was manually reenergized from Bus 28 at 1213 hours; however, LPCI remained inoperable pending investigations.

Restoration of Bus 28/29-5 auto-transfer function occurred at 2151 hours, allowing LPCI to be returned to operable status and TS 3.5.1.E to be exited.

This report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of 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.

05000254/LER-2004-0036 December 2004Quad Cities

On October 8, 2004, at 0300 hours, it was determined that the Control Room Emergency Ventilation System (CREVS) Test had not been performed correctly during the last performance, and TS Surveillance Requirement (SR) 3.0.3 was entered for a missed surveillance.AAt 2120 hours, the surveillance was performed and the pressure differential was determined to be less than required.A The CREVS was declared inoperable, TS SR 3.0.3 was exited and TS Action 3.7.4.A was entered. At 2158 hours, an Emergency Notification System call was made in accordance with 10CFR50.72(b)(3)(v)(D) for loss of safety function of a single train system. The seals for two hatch covers were enhanced and the test was re-performed successfully.AOn October 9, 2004, at 0400 hours CREVS was declared operable.

This event was caused by an inadequate review of a procedure change in 1998, and a deficient modification to the hatch covers in 1999. Corrective actions include enhancements to the seals and a procedure change to add flow criteria to the test procedure.

The safety significance was minimal because CREVS does not impact reactor safety and the Control Room Emergency zone differential pressure was positive at all times.

05000265/LER-2004-00129 March 2004Quad Cities

On January 28, 2004, it was determined during troubleshooting that a wire on the selector switch for the Unit 2 "A" drywell (DW) radiation monitor was crimped back on itself at a connection point but was not soldered. This instrument provides post-accident indication as well as an isolation of Primary Containment (Group II isolation) in response to high radiation levels in the drywell. The troubleshooting was being performed in response to a drop in the indicated value on January 18, 2004, from 3R/hr to 1R/hr.

It was determined that the chassis was manufactured with an unsoldered switch connection, and that this connection made intermittent contact. In the event of a gross failure of the fuel cladding, the discontinuity may have prevented a containment isolation initiation during a DW high radiation condition. However, high DW pressure and low reactor water level instrumentation would have initiated a Group II isolation in the event of a break in the reactor coolant pressure boundary inside containment.

The DW radiation monitors on Unit 1 and Unit 2 were checked and no additional loose or unsoldered connections were identified.

NRC FORM 3 6 6A ( 7 -2 0 01)