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 Report dateSiteEvent description
05000397/LER-2017-00730 November 2017Columbia

On October 3, 2017 at 0800 PDT, Secondary Containment (Reactor Building) became inoperable due to pressure increasing above the Technical Specification (TS) limit of -0.25 inches of water gauge (inwg). While the plant was at 100% power, a Reactor Building exhaust valve unexpectedly closed, resulting in a loss of Secondary Containment for approximately two minutes. Secondary Containment was declared inoperable and TS Action Statement 3.6.4.1.A was entered. The Control Room operators reopened the Reactor Building exhaust valve and pressure returned to within limits automatically. Secondary Containment was declared operable at 0810 PDT and TS Action Statement 3.6.4.1.A was exited. The event was reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D) as Event Notification #52999.

The apparent cause of the event is a surface degradation on the lower stab of an electrical disconnect causing a momentary high resistance when the cubicle door is opened. This event occurred during performance of thermography in the cubicle.

05000397/LER-2017-0069 November 2017Columbia

On September 21, 2017 at 0800 PDT, Columbia Generating Station declared number 2 Diesel Generator inoperable for maintenance and entered Technical Specification Action Statement (TSAS) 3.8.1, Condition B for One Required Diesel Generator inoperable. At 0821 PDT and again at 1537 PDT Surveillance Requirement 3.8.1.1 was performed per action statement B.1. At 2315 PDT it was noted by operations that a page was missing from the surveillance procedure for both occurrences and determined that the action statement was not met within the required action time. Operations entered TSAS 3.8.1, Condition F for Required Action and associated Completion Time of Condition B not being met. AT 2321 PDT SR 3.8.1.1 was performed satisfactorily and Condition F was exited.

The apparent cause was a failure to page check the surveillance procedure prior to use. Initiatives have been developed to reinforce existing standards and improve human performance.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) Operation or Condition Prohibited by Technical Specifications.

05000397/LER-2017-0059 November 2017Columbia

On September 12, 2017 at 1227 PDT, Secondary Containment became inoperable due to pressure increasing above the Technical Specification limit of -0.25 inches of water gauge. While the plant was at 100% power, a Reactor Building exhaust valve and supply valve unexpectedly lost power and closed, resulting in a loss of Secondary Containment for approximately one minute. While Technical Specification limits were exceeded for this short time period, the resulting pressure excursion was bounded by analytical results; and thus, there were no safety consequences for this condition. This event was reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D) as Event Notification #52966.

The apparent cause of the event was that station personnel did not deliberately and conservatively perform work tasks. Workers failed to update work instructions when work was rescheduled, and did not verify power sources at the work site. Corrective actions for this event include conducting a workshop on management expectations of Maintenance, increased management oversight, and addressing human performance issues.

05000397/LER-2017-00417 October 2017Columbia

On August 20, 2017 at 1605 PDT, Columbia Generating Station was manually scrammed due to a rise in Main Condenser back pressure. The rise in back pressure was due to the spurious closure of the Main Condenser Air Removal Suction Valve (AR-V-1) as a result of the failure of it's associated solenoid pilot valve. Following the reactor scram and depressurization of the reactor a Level 3 actuation occurred. In addition a startup flow control valve failed which necessitated throttling of the Feedwater start-up level control isolation valve to control Reactor Pressure Vessel level. All other safety systems functioned as expected and all control rods were fully inserted. Reactor decay heat was removed via bypass valves to the main condenser.

The apparent cause was the plant modification to address the single point vulnerability of the closure of AR-V-1 was not implemented in time to prevent a plant shutdown. A temporary modification has been installed to maintain AR-V-1 open for the remainder of the operating cycle.

These events are reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A).

05000397/LER-2017-00324 August 2017Columbia

On June 6, 2017 at 1756 PDT hours Secondary Containment pressure exceeded the Technical Specification (TS) limit during a period of inclement weather. At 1756 PDT Secondary Containment was declared inoperable and operations personnel entered TS Action Statement 3.6.4. I .A and subsequently exited at 1800 PDT. Secondary Containment pressure was restored automatically by system response and operator action was not required.

The direct cause of the momentary loss of Secondary Containment was due to slow system response to maintain a vacuum in Secondary Containment during a period of inclement weather. The interim planned corrective action is to verify proper operation and tuning of the Secondary Containment instrumentation. Additionally Columbia Generating Station is pursuing the change to TS requirements by adopting TSTF-551, Revise Secondary Containment Surveillance Requirements.

This condition is being reported under 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and to mitigate the consequences of an accident.

05000397/LER-2016-00515 February 2017Columbia

On December 18, 2016, during a forced plant outage reported under Licensee Event Report (LER)-2016-004, a leak was identified on the minimum flow line of the High Pressure Core Spray (HPCS) system downstream of the Primary Containment Isolation Valve.

HPCS system had been running on minimum flow after being used to maintain Reactor Pressure Vessel water level. The HPCS line leak was identified during a walk down by Operations personnel after the HPCS pump had been secured. Due to the location of the leak downstream of the Primary Containment Isolation Valve, this leak constituted a breach of Primary Containment. Both HPCS and Primary Containment were declared inoperable.

The cause of the leak was determined to be from a gasketed flange in the HPCS minimum flow piping. Corrective actions included replacing the gasket. Further evaluation is ongoing and this report will be supplemented once complete.

05000397/LER-2016-00312 January 2017Columbia

On November 20, 2016 at 1402 PST, Secondary Containment (NH) (Reactor Building) became inoperable due to pressure increasing above the Technical Specification limit of -0.25 inches of water gauge (inwg). While the plant was ascending in power, the Reactor Building exhaust air fan unexpectedly failed to start in manual during post-maintenance testing. Prior to this event, Reactor Building Heating, Ventilation and Air Conditioning (VA) (HVAC) System A was running. Per station procedures, System A was stopped and System B was to start. The fan's failure to start resulted in no Reactor Building fans running, and increased Reactor Building pressure.

For a time period of less than one minute, Secondary Containment pressure was not maintained less than or equal to -0.25 inwg.

Immediate recovery actions by Operations personnel included manually starting Reactor Building HVAC System A, which quickly restored Secondary Containment pressure to less than or equal to -0.25 inwg at 1403 PST. While TS limits were exceeded for this short time period, the resulting pressure excursion was bounded by analytical results; and thus, there were no safety consequences for this condition. This event was reported under reporting criteria 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D) as Event Notification #52382.

The cause of the exhaust fan's failure to start was a faulty control switch for the fan. Corrective actions for this event include replacement of the control switch. There were no other event-related equipment malfunctions.

05000397/LER-2016-00125 July 2016Columbia

At 1322 PDT on March 28, 2016, a manual reactor scram was initiated in response to a loss of Reactor Closed Cooling (RCC). The loss of RCC was due to the opening of a Service Water (SW) valve at the inlet side of the Fuel Pool Cooling heat exchanger during performance of a partial surveillance without proper isolation of the RCC system piping from the heat exchanger. The cross-connection of the two systems caused depressurization and loss of flow from the RCC system into the non-pressurized SW piping. The SW valve was closed and the reactor was scrammed. Safety system responses to the scram signal were normal, with all control rods being fully inserted. Reactor decay heat was removed via bypass valves to the Main Condenser. No safety relief valves lifted and no emergency core cooling systems injected following the reactor scram.

The root cause was determined to be that plant Operators did not properly evaluate plant configuration when performing a partial surveillance including the marking as "N/A" (not applicable) of procedural steps, in accordance with plant procedures. Human performance aspects of the event were quickly addressed and additional corrective actions include reinforcing and monitoring procedure standards, and updating work control procedures at the station.

26158 R6 NRC Form 366 (01-2014)

05000397/LER-2015-0077 January 2016Columbia

On November 9, 2015 at 20:40 PST, Secondary Containment (Reactor Building) became inoperable due to pressure increasing above the Technical Specifications (TS) limit of -0.25 inches water gauge (inwg).

At the time of the event the Division 2 Reactor Building Heating, Ventilation and Air Conditioning System (HVAC) was controlling Secondary Containment differential pressure. Power supply E-E/S-299 then failed, causing Division 2 Secondary Containment Pressure controller to lose power. This resulted in the Division 2 Reactor Building Exhaust Fan flow being reduced, causing Secondary containment pressure to rise above TS limit of -0.25 inwg.

Operations personnel manually started the Division 2 SGT lead fan to restore negative pressure. The lead fan operated at max flow (due to the failure of E-E/S-299) resulting in the restoration of Secondary Containment pressure to within TS limits.

The Division 1 HVAC was manually started, allowing Operations personnel to manually secure the Division 2 SGT lead fan and maintain Secondary Containment pressure.

The direct cause for the loss of E-E/S-299 was due to an incorrect lug size installed in the fuse block during initial construction.

Current procedures are adequate to prevent a similar error.

26158 R6 NRC Form 366 (01-2014) APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 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 intemet e-mail to Inf000llects.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.

05000397/LER-2015-00520 August 2015Columbia

On June 25, 2015 it was discovered that Columbia Generating Station's (Columbia) Reactor Protection System (RPS) trip logic was unable to generate a full scram on Reactor Pressure Vessel (RPV) low level because RPS 'A' level 3 indicating switches were mechanically bound high off scale. Immediate actions were taken to comply with Technical Specifications, and a half scram was generated on RPS trip system 'A' to restore full scram capability. Corrective actions include aligning Columbia's maintenance procedures with vendor recommendations, establishing preventative maintenance to ensure correct setting of the indicating switches and verification that level switches are on scale prior to entering the mode of applicability.

26158 R6

05000397/LER-2015-0037 July 2015Columbia

The condition reported by this LER was an expected condition, which was the result of planned activities in support of a routine refueling outage. As described in the LER, the U.S. Nuclear Regulatory Commission (NRC) provided enforcement guidance, applicable to boiling water reactor licensees, that allows the reported condition.

Although this allowance is provided by the NRC's enforcement guidance, the planned activities are still reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications (TS).

Between May 13, 2015 and June 13, 2015, Columbia Generating Station (Columbia) performed Operations with the Potential for Draining the Reactor Vessel (OPDRV) activities while in Mode 5 without an operable secondary containment, as expected and allowed by NRC Enforcement Guidance Memorandum (EGM) 11-003, Revision 2.

Although EGM 11-003, Revision 2, allows implementation of interim actions as an alternative to full compliance, this condition is still considered a condition prohibited by Technical Specification (TS) 3.6.4.1. The OPDRV activities were planned activities that were completed under the guidance of plant procedures and work instructions and are considered to have low safety significance based on the interim actions taken. Since these actions were deliberate, no cause determination was necessary. A license amendment request will be submitted following NRC approval of the Technical Specification Task Force (TSTF) traveler associated with generic resolution of this issue.

26158 R6 NRC Form 366 (01-2014) APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 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 intemet e-mail to Inf000llects.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.

05000397/LER-2014-0049 October 2014Columbia

On August 14, 2014 it was discovered that Columbia Generating Station's (Columbia) method of complying with Technical Specification (TS) Surveillance Requirement (SR) 3.7.1.1 for Ultimate Heat Sink (UHS) spray pond level was Inadequate. The SR requires level In each spray pond to be verified to be greater than or equal to the minimum water level whereas procedures allowed for an arithmetic average of the two ponds to be taken when a single Service Water (SW) pump is in operation which creates a differential between the pond levels. The two spray ponds that make up Columbia's UHS are connected by a siphon line to allow water to be shared between the two ponds. Columbia's original TS did not specify that a minimum water level be checked in each pond and a procedural note was added to clarify compliance with the TS SR during single SW pump operation. When Columbia upgraded the . TSs in 1997 the word 'each' was introduced to the TS. Corrective actions include a TS amendment submitted to the NRC on August 22, 2014.

  • , 26158 Re NRC Form 368 (01-2014) APPROVED BY OMB: NO. 3160-0104 EXPIRES: 01/31/2017 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-6 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20655-0001, or by hornet e-mail to Infocollects.Resourceenrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20603.11 a means used to impose an inlormadon 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.

05000397/LER-2014-00312 May 2014Columbia

FACILITY NAME =

M. Hedges

  • TELEPHONE NUMBER (Include Anon Cod.) 509-377-8277

13. COMPUTE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT

CAUSE SYSTEM COMPONENT MANU- FACTURER

REPORTABLE

TO EPIX CAUSE SYSTEM COMPONENT

MANU-

FACTURER

REPORTABLE

TO EFIX

14. SUPPLEMENTAL REPORT EXPECTED

0 YES (ft yes, complete 15. EXPECTED SUBMISSION DATE) .:. NO 14. EXPEL MONTH DAY r YEAR

SUBMISSION

DAT3 ., .. . 4.

ABSTRACT (Limit to I it spaces, i.e., approximately 15 single-spaced typewritten On March 12, 2014, it was identified that thr: manhole covers (E10, El 1, and E 15) for vaults containing 4160 volt electric cables were missing the hold down bolts. The hold down bolts are required as part of the tornado missile barrier for the manhole cover for the underground electric vault. It was later determined that the hold down bolts for manhole cover E 1 / (Division 2 Service Water) had been identified as missing since September 6, 2013.

When the bolts were identified as missing on September 6, 2013, the manhole cover and bolts were not recognized as a tornado missile barrier; because this information was not available in routinely used databases and procedures. No compensatory action was taken until March 12, 2014, when large concrete blocks were placed on top of the manholes to prevent the covers from potential removal in the event of a tornado. The degraded missile barrier is considered to have rendered the Division 2 Service Water system inoperable from September 6, 2013, to March 12, 2014. This is reportable as a condition prohibited by Technical Specifications.

26158 R6 NRC Form 366 (01-2C14) APPROVED DV ON& NO. 31104104 - UMW: 01l3I/2017 Elrodad Laden par respire lo comply sib Ilia mandelory colecim request P3 hours.

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NRC FO

U.S. NUCLEAR REGULATORY COMMISSION UCENSEE EVENT REPORT (LER)

CONTINUATION SHEET

Columbia Generating Station 05000 397 & LER NUMBER 3 P AG E 2 OF 2014 - 0

PLANT CONDITIONS

Columbia was operating at 100% power. There were no inoperable plant systems at the time of discovery that contributed to this event.

EVENT DESCRIPTION

On March 12, 2014, it was identified that three manhole covers (E10, El 1, and E 15) for vaults containing 4160 volt electric cables (CBI.) were missing the hold down bolts. The electric cables in vaults El0 and Eli support the Division 2 Service Water (BI) system. The electric cables in vaults EIS s the Division 3 Service Water system. The hold down bolts are required as part of the tornado missile barrier for the manhole cover for the underground electric vault. It was later determined that the hold down bolts for manhole cover Ell (Division 2 Service Water) had been identified as missing in two condition reports on September 8, 2013, and again on December 3, 2013. When Operations performed an immediate operability determination for the September and December 2013 condition reports, the manhole cover and bolts were not recognized as a tornado missile barrier; therefore, no compensatory action was taken until March 12, 2014.

The degraded missile barrier is considered to have rendered the Division 2 Service Water system inoperable from September 6, 2013, to March 12, 2014. This is reportable as a condition prohibited by Technical Specifications.

IMMEDIATE CORRECTIVE ACTION

It was verified through the Weather Service that no tornados were predicted for our area in the next 24 hours. Large concrete blocks were placed over the manhole covers on March 12, 2014, to prevent the cover from lifting in the event of a tornado.

The information fields in the master equipment list for the applicable manhole covers were completed identifying them as a tornado missile barrier.

CAUSE

The information in this section is based on the preliminary results of a root cause evaluation. If any significant changes in the cause or corrective actions are made in the final evaluation, a supplement will be submitted for this report.

The direct cause for the missing hold down bolts was not determined.

The root cause of the failure to recognize the manhole cover as a tornado barrier was that station procedures that implement the process to establish quality classifications for safety-related systems, structures, and components (SSCs) did not ensure accurate information was available in a timely manner for these components.

There are multiple methods available for determining the safety significance of a degraded component; however., the computerized master equipment list is typically the preferred method. The equipment plant numbers (EPNs) for the manhole covers had been entered into master equipment list in 2012, but no action was taken to complete the remaining data fields for the components to kientify that the manhole covers fulfilled a tornado barrier function, and have the information verified and approved. The master equipment list that station personnel used identified these manhole covers as nonsafety-related.

26158A A3 NA'; FORM 366A 01-"eC14) IL LER NUMENEPI 5 PAGE 2. DOCKET 1. FACILITY NAME Columbia Generating Station YEAR OF 3 A 0014 - 003 -0

NRC FORM

014014) LICENSEE EVENT REPORT (LER) U.S. NUCLEAR REGULATORY COMMISSION

CONTINUATION SHEET

In the immediate rability determination process for the two Condition Reports in 2013, Operations personnel did not have information readily available to them to determine that the manhole cover was a tornado missile barrier. Most plant personnel, including Operations, did not recognize that the information in the computer master equipment list for the manhole covers was not at an approved status and should not be used. The manhole covers did not have any label indicating that it was a tornado missile barrier. Plant drawings did not identify the manhole covers as tornado missile barriers. Additionally, the station barrier impairment procedure did not list the manhole covers as a tornado missile barrier.

FURTHER CORRECTIVE ACTION

Work requests were initiated to repair/replace the missing bolts for the manhole covers.

Revise Engineering procedures to provide clear direction to establish Quality Classifications for safety-related SSCs within the Master Equipment List within a specific time frame. Ensure that procedures include timeliness requirements for establishing and upgrading/downgrading EPNs for installed plant equipment.

Apply a marking on the manhole covers that identify these covers as a tornado missile barrier.

Revise the station barrier impairment procedure to identify manhole covers as a tornado missile barrier.

ASSESSMENT OF SAFETY CONSEQUENCES

No actual tornados occurred during the time of interest. The Division 2 Service Water system remained capable of fulfilling its safety function during this time period. Additionally, at least one other division of Service Water was available during this time period (September 6, 2013 to March 12, 2014) to be able to fulfill the safety function; therefore, the actual safety consequence of this issue was minimal.

SIMILAR EVENTS

There have been no similar events at Columbia Generating Station in the last three years.

ENERGY INDUSTRY IDENTIFICATION SYSTEM (ENS) INFORMATION CODES EIIS s a are bracketed 'Mere applicable in the narrative.

NERGY

.!‘‘ORT1:1--"T er 914)

05000397/LER-2013-0031 August 2013Columbia

On June 3, 2013, with Columbia Generating Station in a planned refueling outage and the reactor cavity flooded up (Mode 5), leakage past a closed isolation valve associated with one hydraulic control unit on the control rod drive system was observed through a drain line. This leakage originated from the reactor vessel and constituted an operation with the potential to drain the reactor vessel (OPDRV).

When the leakage was initially identified, it was not recognized as an OPDRV. The leakage rate was estimated to be less than 10 gallons per hour. This leakage persisted for 16 hours until the maintenance activities were completed. During this time period, the secondary containment was inoperable. Technical Specifications (TS) require that with secondary containment inoperable during OPDRV activities, action must be initiated immediately to suspend OPDRVs. Contrary to this requirement, action was not taken to suspend the OPDRV. This represents a condition prohibited by TS. The cause of this event was inadequate procedure guidance for actions to take when unexpected OPDRV conditions are encountered. Immediate corrective actions were taken to establish expectations regarding the appropriate actions to take for discovered unplanned OPDRV conditions.

This event is not safety significant since the leakage rate was so small that there was no measurable loss of level in the reactor cavity.

26158 R5

05000397/LER-2013-0018 March 2013Columbia

On January 7, 2013, Columbia was operating at 100% power. At 1427 PST both doors of the 501 elevation airlock entrance of the Reactor Building were simultaneously opened for a short period of time. This was the result of the failure of the interlock between the outer door (R304), and the inner door (R305).

Maintenance personnel were moving scaffolding through the Reactor Building outer security door when an equipment operator opened the inner door to exit the Reactor Building. This equipment operator immediately closed the inner door and contacted the Main Control Room. The outer door was key locked closed until corrective actions could be implemented.

Having a condition where both doors in a Reactor Building airlock were opened simultaneously while not undergoing a planned evolution for maintenance or surveillance testing, results in an unintended entry into Technical Specification (TS) 3.6.4.1 Secondary Containment due to a failure to satisfy Surveillance Requirement (SR) 3.6.4.1.3.

It was determined that this condition resulted in a loss of safety function (Secondary Containment) and was reported to the NRC (Event Notification 48656) at 1629 PST on January 7, 2013.

2615B R5 Ur.

  • 4-';
05000397/LER-2011-00225 October 2011Columbia

At 2021 hours on August 27, 2011, a loss of shutdown cooling occurred due to a spurious undervoltage signal in one of two, in series, Electrical Protection Assembly (EPA) circuit breaker supplying the B train of the Reactor Protection System (RPS) power bus (RPS-B). Response to the spurious signal resulted in loss of power to RPS-B and associated actuations including isolation of the common shutdown cooling suction valves.

The spurious signal originated in a logic board (GE Model 147D8652G007) associated with the EPA Breaker.

Post event testing was unable to specifically identify the discrete component responsible for the failure. The root cause was that Energy Northwest was not proactive in replacing older, obsolete model boards (including the one that caused the event) with a new model recommended by the vendor. The faulty logic board and the other logic board in series for RPS-B were replaced with newer model boards. Further corrective actions will replace the remaining logic boards currently installed in the plant with the newer models. This event is being reported under 10 CFR 50.73(a)(2)(v) as an event that could have prevented fulfillment of a safety function, as well as an invalid actuation of containment isolation in multiple systems per 10 CFR 50.73(a)(2)(iv).

26158 R5 U.S. NUCLEAR REGULATORY COMMISSION

05000397/LER-2011-00129 August 2011Columbia

On June 28, 2011, while the plant was in Mode 5 for refueling outage R20, Columbia Generating Station (Columbia) failed to enter a required Technical Specifications (TS) Action Statement while performing control rod exercises. During stroke time testing, control rod 34-47 displayed an erroneous indication. Upon initial withdrawal, the four rod display showed an alternating indication of "XX" (meaning the reed switch was not open during movement) and "00" (full in indication) requiring the position indication to be declared inoperable per TS 3.9.4. Control rod 34-47 was subsequently fully inserted and testing resumed on other rods contrary to the required action statement of TS 3.9.4. Upon discovery of the noncompliance, the TS required actions were subsequently performed and the failed reed switch replaced. The Control Room Supervisor and Shift Manager did not verify the required action statements specified in the TS and Bases as required. This was determined to be the apparent cause. A contributing cause included not performing all of the required steps in the procedure for control rod stroke time testing. Columbia has had no previous occurrences of a failure to enter the required action statement of TS 3.9.4.

This condition is reportable under 50.73(a)(2)(i)(B) as a condition prohibited by TS.

26158 R5

05000397/LER-2008-00116 March 2010Columbia

65% power due to the Digital Electrohydraulic Control (DEH) trip header momentarily depressurizing during post maintenance testing (PMT). All safety systems were available during the event and operated as designed. Plant operators effectively managed the transient. This event did not pose a threat to the health and safety of the public.

The direct cause of the reactor scram was instantaneous recompression of an air bubble trapped in the intervalve cavity between the A and B Quadvoter valves during post-maintenance testing (PMT).

This allowed backflow of DEH fluid from the emergency trip header into the intervalve cavity during PMT, causing a momentary depressurization of the DEH trip header that resulted in the reactor scram. The root cause of the scram was determined to be a design deficiency in the on-line serviceable Quadvoter assembly which allowed an air bubble to remain in the intervalve cavity following performance of on-line maintenance activities.

26158 R4 Columbia Generating Station 05000397

05000397/LER-2009-00316 March 2010Columbia

the #2 high pressure turbine bearing was reported to the main control room. As directed by procedure, a manual scram was inserted at 1952.

The in-service turbine lube oil vapor extractor system did not provide enough differential pressure to prevent lube oil from leaking out of the bearing pedestal. The fire occurred when the leaking lube oil came into contact with a hot pipe causing the oil to flash.

The root causes were an out of calibration pressure switch and the integrated system knowledge of the main turbine lube oil exhauster system by Operations and Engineering was weak. The pressure switch has been replaced. Additional corrective actions taken to prevent recurrence include revisions to station procedures to ensure that the turbine lube oil vapor extractor system pressure is monitored via independent instruments during routine operator rounds as well as after any system manipulation. In addition, improvements in training fOr Operations and Engineering are planned to increase overall system knowledge.

No similar events have been reported by Energy Northwest.

26158 R4 Columbia Generating Station

05000397/LER-2007-0057 February 2008Columbia

On December 10, 2007, it was discovered that an unidentified failure of the Emergency Diesel Generator (DG) that supports the High Pressure Core Spray system resulted in a failure to comply with the required actions of three separate conditions of Technical Specification 3.8.1, AC Operating Sources on two separate occasions. The cause of the DG failures was the performance of inadequate procedures on May 3, 2005 and October 19, 2007 that resulted in clearing of the fuses on the primary side of the metering and relaying potential transformers during shut down of the DG. The potential transformers provide power to the electronic governor as well as the local and remote indications rendering the electronic governor inoperable while the fuses were cleared. The DG was inoperable from May 3, 2005 to June 7, 2005 and again from October 19, 2007 until November 10, 2007.

The root cause of the inadequate procedures was a lack of knowledge of the DG shut down logic by licensee Operations and Engineering personnel. Corrective actions include revising the affected procedures and providing training for the appropriate Operations and Engineering personnel.

This event did not adversely affect the health and safety of the public because the DG remained available and no loss of off-site power occurred during the time frames the fuses were cleared.

26158 R3

05000397/LER-2004-0044 October 2004Columbia

On July 30, 2004, Columbia Generating Station (Columbia) was in Mode 1 with the reactor operating at -100 percent power. At 09:23 PDT, the reactor automatically scrammed when the reactor protection system (RPS) received trip signals from three out of four reactor steam dome pressure - high instrument channels.

The high reactor steam dome high-pressure condition was a result of a turbine governor valve (MS-V­ GV/1) drifting closed. The turbine governor valve drifted closed due to a failure of a bypass capacitor on a NUCANA Servo Driver (NSD) circuit board associated with the governor valve electro-hydraulic control system. The failed capacitor was a monolithic ceramic capacitor. This capacitor provides high frequency bypass filtering for the onboard power supply at one of the operational amplifiers. The capacitor failed with low resistance which caused a high current load that eventually caused the circuit board protective fuse to clear and the closure of MS-V-GV/1.

This event posed no threat to the health and safety of the public. The NSD circuit board was replaced and a detailed failure analysis will be performed on the failed circuit board.

26158 R2

05000397/LER-2004-00123 March 2004Columbia

At 1315 on January 23, 2004, with the plant in Mode 1 at approximately 100 percent rated thermal power, it was determined that a condition prohibited by the Columbia Generating Station Technical Specifications (TS) existed from 1500 on October 27 through 2100 on October 29, 2003 and from 1305 on November 10 through 1052 on November 11, 2003 when testing was conducted to measure control room in-leakage. During this test, common ducts for both Control Room Emergency Filtration (CREF) __subsystems werebreached multiple times by removing duct access panels to install and remove test equipment. During the time the access panels- v`iere removed, floWithii10 CFR -eTeated -thatVolild-have challenged the CREF's ability to pressurize the control room and mitigate the consequences of an accident. Both CREF subsystems were determined to have been inoperable when the access panels were removed. With two CREF subsystems inoperable, TS 3.7.3 directs immediate entry into TS Limiting Condition for Operation (LCO) 3.0.3. The fact that removing the access panels caused both CREF subsystems to become inoperable was not recognized at the time the tests were performed because the HVAC ducts were allowed to be breached for short periods of time by the plant barrier impairments procedure.

The cause of this event is attributed to inadequate guidance in the Columbia Generating Station barrier impairment procedure. The cause of the inadequate guidance in the procedure was a lack of understanding of regulatory guidance associated with barrier impairments.

26158 RI