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 Start dateReporting criterionTitleEvent descriptionSystemLER
ENS 5411312 June 2019 03:04:0010 CFR 26.719, FFD Reporting requirementsFitness for Duty - Contract Employee Supervisor Confirmed Positive for AlcoholA contract employee supervisor had a confirmed positive test for alcohol during a random fitness-for-duty test. The employee's access to the plant had been terminated. The NRC Resident Inspector has been notified.
ENS 538604 February 2019 22:03:0010 CFR 26.719, FFD Reporting requirementsFitness for Duty - Licensed Operator Confirmed Positive for Drug TestA licensed reactor operator had a confirmed positive random fitness-for-duty drug test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 537754 December 2018 08:00:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessLoss of Emergency Assessment CapabilityOn 12/4/2018 at 1340 (PST), Columbia entered a planned evolution to replace the seismic monitoring system. Use of the Modified Mercalli Intensity Scale has been implemented as a compensatory measure per station procedures. The expected duration of the replacement activity will exceed 72 hours, therefore, this is being reported as a major loss of emergency assessment capability in accordance with regulation 10 CFR 50.72(b)(3)(xiii). Compensatory measures will remain in place until the seismic system replacement has been completed. The NRC Resident Inspector has been notified."
ENS 5341018 May 2018 13:51:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram Caused by Main Transformer TripAt 0651 (PDT) on May 18th, 2018, Columbia Generating station experienced a Main Transformer trip, that caused a Reactor Scram. Reactor Power, Pressure and Level were maintained as expected for this condition. MS-RV-1A (Safety Relief Valve) and MS-RV-1B (Safety Relief Valve) opened on reactor high pressure during the initial transient. MS-RV-1B appeared to remain open after pressure lowered below the reset point. The operating crew removed power supply fuses for MS-RV-1B and it currently indicates intermediate position. SRV (Safety Relief Valve) tail pipe temperatures indicate all valves are closed. Suppression pool level and temperature have remained steady within normal operating levels. All control rods inserted and reactor power is being maintained subcritical. RPV (Reactor Pressure Vessel) water level is being maintained with condensate and feed system with startup flow control valves in automatic. Reactor Pressure is being maintained with the Turbine Bypass valves controlling in automatic. The main condenser is the heat sink. No ECCS (Emergency Core Cooling Systems) systems actuated or injected; the EOC-RPT (End of Cycle-Recirculation Pump Trip) and RPS (Reactor Protection System) systems actuated causing a trip of the RRC pumps and a reactor scram. Core recirculation is being maintained with RRC-P-1A (Reactor Recirculation Pump) running. No release has occurred. At this time there will be no notifications to state, local or other public agencies. The NRC Senior Resident has been notified. The cause of the event is currently under investigation. Plant conditions are stable. The plant is in its normal electrical alignment and offsite power is available to the site.Reactor Pressure Vessel
Reactor Protection System
Main Transformer
ENS 5306813 November 2017 16:02:0010 CFR 26.719, FFD Reporting requirementsFitness for Duty - Positive Result for AlcoholA licensed supervisor had a confirmed positive test for alcohol. The employee was escorted offsite and their plant access has been terminated and a five year denial placed in Personnel Access Data System (PADS). This is being reported per 10 CFR 26.719(b)(2)(ii). The NRC Resident Inspector has been notified.
ENS 529993 October 2017 15:00:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Secondary Containment Inoperable Due to Unexpected Isolation of Exhaust ValveOn October 3, 2017, at 0800 PDT, Reactor Building (Secondary Containment) pressure momentarily rose above the Technical Specification (TS) limit. Secondary Containment was declared inoperable and TS Action Statement 3.6.4.1.A was entered. The pressure rise was due to unexpected isolation of an exhaust valve in the Reactor Building ventilation system during electrical switchgear inspections. The cause of the closure is still under investigation. The Control Room operators reopened the Reactor Building exhaust valve and pressure returned to within limits automatically. Secondary Containment was declared operable at 0802 PDT and TS Action Statement 3.6.4.1.A was exited. This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(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 accident mitigation. The NRC Resident Inspector has been notified.Secondary containment
Reactor Building Ventilation
05000397/LER-2017-007
ENS 5296612 September 2017 19:28:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Secondary Containment Pressure Momentarily Above Technical Specification LimitOn September 12, 2017, at 1228 PDT, Reactor Building (Secondary Containment) pressure momentarily rose above the Technical Specification (TS) limit. Secondary Containment was declared inoperable and TS Action Statement 3.6.4.1.A was entered. The pressure rise was due to unexpected isolation of the supply and exhaust valves in the Reactor Building ventilation system due to an electrical transient on the power panel feeding the valve operators' solenoid pilot valves during maintenance. The cause of the electrical transient is under investigation. The Reactor Building differential pressure controller restored the building pressure to within limits. The Control Room operators reopened the Reactor Building ventilation supply and exhaust valves. Secondary Containment was declared operable at 1228 PDT and TS Action Statement 3.6.4.1.A was exited. This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(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 accident mitigation. The licensee notified the NRC Resident Inspector.Secondary containment
Reactor Building Ventilation
05000397/LER-2017-005
ENS 5291820 August 2017 23:05:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Scram Due to a Rise in Main Condenser Back Pressure

On August 20, 2017 at 1605 PDT, Columbia Generating Station was manually scrammed from 100 percent power due to a rise of Main Condenser back pressure. Manual scram of the unit is procedurally required upon a loss of Main Condenser back pressure. Preliminary investigations indicate that the Main Condenser air removal suction valve (AR-V-1) closed, resulting in the Condenser back pressure rising to within 1.0 inch Hg of the setpoint with reactor power greater than 25 percent. Further investigations continue. All control rods fully inserted. In addition to the closure of the air removal suction valve, one of two Reactor Feedwater startup flow control valves did not adequately operate to control Reactor vessel level and resulted in a high-level (Level-8) actuation tripping the Reactor Feedwater System. All other systems operated as expected. Reactor water level is currently being controlled manually with the start-up level control isolation valve. AR-V-1 has been manually opened with a jumper and temporary air supply. Reactor decay heat is being removed via bypass valves to the Main Condenser. This event is being reported under the following: 10 CFR 50.72(b)(2)(iv)(B), which requires a four-hour notification for any event or condition that results in actuation of the Reactor Protection System when the reactor is critical. The licensee notified the NRC Resident Inspector. The licensee plans to issue a press release.

  • * * UPDATE ON 8/24/17 AT 1937 EDT FROM MATT HUMMER TO DONG PARK * * *

The licensee is updating the notification to include an 8 hour notification under 10 CFR 50.72(b)(3)(iv)(A) for a specified system actuation due to a Level 3 isolation signal which occurred approximately 20 minutes after the scram. The licensee is currently in cold shutdown to repair the Reactor Feedwater startup flow control valve. The licensee has notified the NRC Resident Inspector. Notified R4DO (Farnholtz).

Feedwater
Reactor Protection System
05000397/LER-2017-004
ENS 5291312 August 2017 07:00:0010 CFR 21.21(d)(3)(i), Failure to Comply or DefectPart 21 Report - Relay Mounting Plate Mounted Upside DownPursuant to 10 CFR 21, this is a non-emergency notification by Energy Northwest concerning a defect in General Electric (GE) Nuclear HMA124A2 relays received at Columbia Generation Station. On August 12, 2017, Energy Northwest completed a 10 CFR 21 evaluation of a condition associated with GE Nuclear HMA124A2 relays supplied by GE Hitachi Nuclear Energy Americas, LLC, and intended for use at Columbia Generating Station. The evaluation was performed to determine the applications where the relays were approved for installation, and where they were installed in the plant, and to determine if the failure of the relays could result in a Substantial Safety Hazard as defined In 10 CFR 21.3. Two of the HMA124A2 relays received had back plates that were mounted upside down, causing the terminals to not match the standard configuration. Although the internal wiring to the physical stud locations was correct, the numbering scheme embossed on the back plate did not match the correct configuration. With the incorrectly mounted back plate, the internal coil of the energizing circuit could be wired to the incorrect portion of the control circuitry, which would not energize when required and could result in the failure of a safety function. This deviation presents a Substantial Safety Hazard as defined In 10 CFR 21.3, as these relays were approved for use in safety related applications; however, there was no actual risk to plant safety since this deviation was recognized and resolved by station craft prior to installation of the relays. This condition is reportable under 10 CFR 21.21(d)(1) as a defect as defined in 10 CFR 21.3. The defective HMA124A2 relays were installed in the plant in the correct configuration with post-maintenance testing performed to ensure operability of the relays. The remaining HMA124A2 relays were examined and no additional defects were identified. GE Hitachi has been notified of the condition. The licensee will notify the NRC Resident Inspector.
ENS 5284811 July 2017 14:50:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentFlow Indicating Switch for High Pressure Core Spray Unreliable Indication

This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented fulfillment of a safety function. On July 11th, 2017, it was discovered that the flow indicating switch for the high pressure core spray (HPCS) minimum flow valve was providing unreliable indication. There was no flow through the line at the time the condition was discovered. This switch provides the flow signal to the HPCS minimum flow valve logic. The switch was declared inoperable and the required actions of Technical Specification 3.3.5.1 were entered. This condition could have prevented the HPCS system, a single train safety system, from performing its specified safety function. Troubleshooting is underway to determine the cause of and correct the condition. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION FROM DAN SHARPE TO KARL DIEDERICH AT 1710 EDT ON 9/20/17 * * *

The condition reported in Event notification #52848 pursuant to 10 CFR 50.72(b)(3)(v)(D) has been evaluated, and determined not to have met the threshold for classification as an Event or Condition the Could Have Prevented Fulfillment of a Safety Function. Engineering analysis has concluded that the affected switch was capable of performing its required support function to provide the flow signal to the HPCS minimum flow valve logic. Thus, the HPCS system remained capable of performing its specific function for the identified condition. The NRC Resident Inspector has been notified. Notified R4DO (G. Miller).

High Pressure Core Spray
ENS 5283027 June 2017 00:56:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Secondary Containment Declared InoperableOn June 26, 2017 at 1756 PDT, Reactor Building (Secondary Containment) pressure rose above the Technical Specification (TS) requirement multiple times from 1756 to 1800. Secondary Containment was declared inoperable and entry into Technical Specification Action Statement 3.6.4.1.A was made. There was a significant change in average wind speed and barometric pressure occurring at that time. At 1800 PDT, Secondary Containment pressure was restored to within limits and TS 3.6.4.1.A was exited. Environmental conditions have stabilized. The Reactor Building Differential Pressure controllers are working as designed. This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(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 accident mitigation. The licensee has notified the NRC Resident Inspector.Secondary containment
ENS 5286925 May 2017 23:47:0010 CFR 50.73(a)(1), Submit an LERInvalid High Pressure Core Spray (Hpcs) System Actuation in Mode 5At 1647 (PDT) on May 25, 2017, during the performance of a post-maintenance test for replacement of a Reactor Pressure Vessel (RPV) low water level 3 indicator switch (MS-LIS-24A), a pressure perturbation in the common pressure reference line resulted in tripping of the RPV Level 2 instruments and an unplanned start of High Pressure Core Spray (HPCS) pump (HPCS-P-1) and its supporting emergency diesel (DG3). The Reactor Pressure Vessel was flooded up during the refueling outage, thus, the actuations of the HPCS pump and its supporting emergency diesel (DG3) were unplanned and invalid. The HPCS pump did not inject into the RPV due to the RPV level being above Level 8 which is an interlock to close the HPCS RPV injection valve (HPCS-V-4). During the event, the single train HPCS system initiated normally but did not inject into the reactor pressure vessel as expected due to flooded-up conditions of the reactor pressure vessel for refueling outage activities. The emergency diesel generator started normally in response to the initiation signal of HPCS. Both HPCS and the emergency diesel generator functioned successfully. All systems responded in conformance with their design and there was no safety significance associated with this event. At the time of the event, the licensee notified the NRC Resident (Inspector).High Pressure Core Spray
Reactor Pressure Vessel
Emergency Diesel Generator
ENS 5261514 March 2017 23:00:0010 CFR 21.21(d)(3)(i), Failure to Comply or DefectPart 21 Notification - Evaluation of a Deviation - Starter Contactor FailurePursuant to 10 CFR Part 21, this is a non-emergency notification by Energy Northwest concerning a defect in Size 1 Freedom Series Starters with nominal 120 VAC coils manufactured by AZZ/NLI (Nuclear Logistics Inc.) used at Columbia Generating Station. On February 8, 2017 Energy Northwest was notified by NLI of a deviation associated with starter contactors used at Columbia that failed to close due to overheating of the starter coil. The coils that were provided were determined to not meet specified voltage ratings. The evaluation completed by Energy Northwest on March 14, 2017 concluded that the deviation did create a Substantial Safety Hazard, and is reportable under 10 CFR 21.21(d)(1) as a defect. A 30 day report will be issued by April 13, 2017 per 10 CFR 21.21(d)(4). The licensee performed a prompt operability assessment for the two starters currently installed. The licensee will notify the NRC Resident Inspector.
ENS 5251026 January 2017 02:36:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentHigh Pressure Core Spray System Diesel Room Fan Motor FailureOn January 25, 2017, at 1836 PST, smoke was detected in the High Pressure Core Spray System (HPCS) diesel room with no indication of a fire. Investigation found the motor starter coil for DMA-FN-32 (Diesel Mixed Air Fan 32), HPCS diesel generator room normal cooling fan, failed. This fan is required for operability of the switchgear that powers the HPCS pump. The HPCS pump is currently inoperable due to maintenance being performed on other support systems. This condition is being reported under 10 CFR 50.72(b)(3)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector.High Pressure Core Spray05000397/LER-2017-001
ENS 5244319 December 2016 07:20:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
Unisolable Leak on High Pressure Core Spray

On December 18, 2016 at 2320 (PST), a leak was discovered on the High Pressure Core Spray (HPCS) system minimum flow line. The leak is located at a bolted flange downstream of the manual isolation valve HPCS-V-53. The location of the leak is not isolable from the suppression pool. This provides a direct path from inside the Primary Containment to the Reactor Building. High Pressure Core Spray system is a single train Emergency Core Cooling System (ECCS) system, therefore inoperability is reportable per 10 CFR 50.72(b)(3)(v)(D). Based on the location of the leak, Primary Containment integrity is compromised. Primary Containment was declared inoperable and is reportable per 10 CFR 50.72(b)(3)(ii)(A). The cause of the leak is under investigation. Actions are underway to cool down and enter MODE 4. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM MATT HUMMER TO HOWIE CROUCH AT 2245 EDT ON 5/24/17 * * *

Engineering evaluations indicate that there was neither a High Pressure Core Spray (HPCS) system inoperability nor a condition that resulted in a significantly degraded principal safety barrier (Primary Containment). Therefore, this event does not meet the reporting criteria in 10 CFR 50.72(b)(3)(v)(D) and 10 CFR 50.72(b)(3)(ii)(A), and Event Notification# 52443 is being retracted. Bases for the retraction are: (1) Extent or accumulation of water flooding the HPCS room would not have prevented the system from fulfilling any of its designated safety functions, if the system had received a starting signal due to an emergency; and (2) the consequences of the HPCS Minimum Flow Line leak into the Reactor Building were within the dose limits and did not have a significant effect on Primary Containment integrity; therefore, the Primary Containment was degraded but operable. The licensee has notified the NRC Resident Inspector. Notified R4DO (Groom).

High Pressure Core Spray
Primary containment
05000397/LER-2016-005
ENS 5244218 December 2016 19:24:0010 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Scram Due to Load Reject from SubstationOn December 18, 2016 at time 1124 PST the plant experienced a full reactor scram. Preliminary investigations indicate that the scram was caused by a load reject from the Bonneville Power Administration (BPA) Ashe substation. Further investigations continue. The following conditions have occurred: Turbine Governor valve closure Reactor high pressure trip +13 inches reactor water level activations E-TR-B (backup transformer) supplying E-SM-7/SM-8 (vital power electrical busses) Complete loss of Reactor Closed Cooling (RCC) E-TR-S (Startup transformer) supplying SM-1/2/3 (non-vital power electrical busses) E-DG-1/2/3 (emergency diesel generators) auto start Low Pressure Core Spray (LPCS) and Residual Heat Removal (RHR) A/B/C initiation signals Main Steam Isolation Valves (MSIV) are closed Reactor Core Isolation Cooling (RCIC) RCIC and High Pressure Core Spray (HPCS) were manually activated and utilized to inject and maintain reactor water level. Pressure control is with Safety Relief Valves (SRV) in, manual. Level control is with RCIC and Control Rod Drive (CRD). RCIC has experienced an over speed trip that was reset so that level control could be maintained by RCIC. This event is being reported under the following: 10 CFR 50.72(b)(2)(iv)(A) which requires a 4 hour notification for Emergency Core Cooling System (ECCS) discharge into the reactor coolant system. 10 CFR 50.72(b)(2)(iv)(B) which requires a 4 hour notification for any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical. 10 CFR 50.72(b)(3)(iv)(A) which requires an 8 hours notification for actuation of ECCS systems. All control rods fully inserted. The NRC Resident Inspector has been informed. The licensee indicated that no increase in radiation levels were detected.Emergency Diesel Generator
Core Spray
Residual Heat Removal
Main Steam Isolation Valve
Reactor Core Isolation Cooling
High Pressure Core Spray
Reactor Coolant System
Reactor Protection System
05000397/LER-2016-005
ENS 5238220 November 2016 22:02:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Secondary Containment Differential Pressure Less than Technical Specification RequirementOn November 20, 2016 at 1402 PST, Reactor Building Exhaust Air Fan 1B, REA-FN-1B, failed to start in manual which caused the Technical Specification (TS) for secondary containment pressure boundary to not be met. The duration of the time that the secondary containment TS was not met was approximately less than one minute. REA-FN-1B was being started in manual during a shift of Reactor Building Ventilation to support a post-maintenance support task on REA-FN-1B. Secondary containment differential pressure was restored within the TS requirement of greater than or equal to 0.25 inch of vacuum water gauge by restarting Reactor Building HVAC Train A. The cause of REA-FN-1B failing to start is currently under investigation. This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(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 accident mitigation. The licensee has notified the NRC Resident Inspector.Secondary containment
Reactor Building Ventilation
HVAC
05000397/LER-2016-003
ENS 522763 October 2016 17:08:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive MaterialLoss of Secondary Containment Vacuum for Four MinutesOn October 3, 2016, at 1008 PDT a Reactor Building Exhaust Valve (REA-V-1) unexpectedly closed, which caused the Technical Specification (TS) for secondary containment pressure boundary to not be met. The duration of time that the secondary containment TS was not met was approximately 4 minutes. Secondary containment differential pressure was restored within TS requirement of greater than or equal to 0.25 inches of vacuum water gauge at approximately 1012 PDT by manually starting Standby Gas Treatment (SGT) system (SYS) A. The cause of the REA-V-1 closure is currently under investigation. This condition is being reported under 10CFR50.72(b)(3)(v)(C) for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector.Secondary containment05000397/LER-2016-002
ENS 5182628 March 2016 20:22:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Scram Following Loss of Reactor Closed CoolingAt 1322 PDT on Monday, March 28, 2016, Columbia Generating Station was manually scrammed from 100% thermal power due to the loss of Reactor Closed Cooling (RCC). Manual scram of the unit is procedurally required upon loss of RCC. The cause of the loss of RCC is being investigated. Regulation 10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours of any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical. All control rods were fully inserted. Valve RWCU-V-4 automatically closed upon high water temperature due to loss of RCC flow. No other safety system actuations were reported. All systems operated as expected. Reactor decay heat is being removed via bypass valves to the Main Condenser. The station is in normal shutdown electrical lineup. The NRC Resident Inspector has been informed. No safety/relief valves lifted and no emergency core cooling systems injected following the reactor scram.Reactor Protection System05000397/LER-2016-001
ENS 5173313 February 2016 18:00:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessRadwaste Building Noble Gas Monitor Out-Of-ServiceOn 2/10/2016 at 1000 PST, Columbia entered a planned evolution to perform channel functional tests on the RadWaste Building noble gas monitor (WEA-RIS-14). Compensatory measures were implemented per station procedures. The station is experiencing equipment issues and the monitor has not been restored within 72 hours (2/13/2016 at 1000 PST) from the start of the outage. The extended outage of this radiological monitoring instrument is, therefore, being reported as a major loss of radiological assessment capability in accordance with regulation 10 CFR 50.72(b)(3)(xiii). Compensatory measures will remain in place until the WEA instrument is restored. The NRC Resident Inspector has been notified. Note: Reactor Power is 75 percent due to a planned plant downpower for unrelated scheduled work with a planned return to 100 percent at 1900 PST on 2/14/16.
ENS 5156224 November 2015 21:30:0010 CFR 50.72(b)(2)(xi), Notification to Government Agency or News ReleaseOffsite Notification of Fuel DefectsAt approximately 1100 PST, Columbia Generating Station (CGS) planned to make a non-required notification to Energy Facility Site Evaluation Council (EFSEC) regarding indications of two fuel defects. This condition has not affected full power operation at CGS, and there is no impact to the health and safety of the public or to the environment. CGS plans on making this notification to EFSEC on November 24, 2015 at 1330 PST. This condition is being reported pursuant to 10 CFR 50.72 (b)(2)(xi). The licensee notified the NRC Resident Inspector.
ENS 5152610 November 2015 04:40:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Reactor Building Vacuum Less than Technical Specifications RequirementAt 2040 PST on 11/9/2015, Reactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge for approximately seven minutes. Operators took action to manually start Standby Gas Treatment System to restore Reactor Building pressure. This event is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. The cause of the event is under investigation. There were no radiological releases associated with the event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector.Secondary containment
Standby Gas Treatment System
05000397/LER-2015-007
ENS 5122814 July 2015 06:39:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Secondary Containment Pressure Increase Above Technical Specification Limit

Reactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge for approximately 2 minutes during a planned surveillance test due to a subsequent failure of REA-FN-1A (Exhaust Fan) to manually start during restoration from the surveillance test. This event is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. Prior to taking test data the surveillance test directs declaring Secondary Containment inoperable in anticipation of potentially exceeding 0.25 inches vacuum water gauge reactor building pressure during the conduct of the surveillance. Consequently Technical Specification LCO 3.6.4.1.A was entered with a 4 hour completion time to restore Secondary Containment to an operable state. Upon failure of REA-FN-1A to start immediate actions were taken to close reactor building ventilation dampers and secure ROA-FN-1A (Supply Fan). Following closure of ventilation dampers and stopping ROA-FN-1A reactor building pressure was quickly restored to less than 0.25 inches vacuum water gauge with Standby Gas Treatment that was already in operation as part of the surveillance test. There were no radiological releases associated with the event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector. Maximum Secondary Containment pressure noted was 0.1 inches positive water gage.

  • * * RETRACTION AT 1351 EDT ON 8/25/2015 FROM MATT HUMMER TO MARK ABRAMOVITZ * * *

Subsequent to the initial report, Columbia has since determined that per NUREG-1022 3.2.7 the event was not reportable as Secondary Containment was 'declared inoperable as a part of a planned evolution ... in accordance with an approved procedure and (Columbia's) TS (Technical Specifications).' No condition has been discovered that would have resulted in the system being declared inoperable prior to the surveillance. Therefore, this event is not considered to be a condition that could have prevented fulfillment of a safety function or a condition prohibited by TS and is not reportable to the NRC as a Licensee Event Report (LEA) per 10 CFR 50.73. The NRC Senior Resident Inspector will be notified. Notified the R4DO (Campbell).

Secondary containment
Reactor Building Ventilation
ENS 512016 July 2015 21:20:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionPostulated Multiple Spurious Operations Scenario That Could Adversely Impact Post-Fire Safe ShutdownA recent review of Fire Protection and Post Fire Safe Shutdown (PFSS) Programs at Columbia Generating Station (CGS) identified a potential unanalyzed condition with Multiple Spurious Operation (MSO) Scenario 2x. Review of the circuit design for High Pressure Core Spray (HPCS) HPCS-V-10, HPCS-V-11 and HPCS-V-15 identified that fire-induced circuit failure (hot shorts) on the OPEN function control circuits for each valve would create the flow path to potentially drain inventory from the suppression pool (SP). The normal operation of HPCS-P-3 (keep-fill pump) would allow additional inventory from the SP to be transferred to the CSTs (Condensate Storage Tank). If a fourth hot short is postulated, HPCS-P-1 would transfer inventory from the SP to the CST at a much faster rate. HPCS-V-11 was deactivated on 6/12/2015 due to a maintenance repair issue and will be left in the fully closed position. This plant alignment resolves current concern for MSO scenario 2x as fire-induced circuit damage cannot cause spurious opening of HPCS-V-11. However, with an incomplete analysis for MSO scenario 2x, compliance with PFSS MSO requirements would have been challenged from the completion of the MSO project (October 2012) up to June 2015. CGS is reporting this event as an unanalyzed condition in conformance with 10 CFR 50.72(b)(3)(ii)(B). Further analyses are being implemented to confirm the condition and to develop appropriate remedial actions. The licensee will notify the NRC Resident Inspector.High Pressure Core Spray05000397/LER-2015-006
ENS 5118226 June 2015 05:00:0010 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the ReactorTwo Reactor Vessel Level Channels Failed HighAt 2200 PDT during startup from refueling outage 22, it was discovered that both level instruments used in reactor protection system (RPS) trip system 'A' for initiation of a reactor scram on low reactor pressure vessel (RPV) level were observed to have failed high. This resulted in the inability to generate a full reactor scram on low level (+13 inches). All remaining RPV level indications demonstrated that level was being maintained within normal operating bands. This constitutes a condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to shut down the reactor. The RPS trip logic at Columbia consists of two trip systems, RPS trip system 'A' and RPS trip system 'B'. There are two level instrument channels in each trip system. Columbia utilizes a 'one-out-of-two taken-twice' trip logic to generate a full scram signal. At least one channel in both trip systems must actuate to generate a full scram signal. With both level instruments in RPS system 'A' failed high, the RPS trip logic was unable to generate a full scram. At 2246 (PDT) and in accordance with TS LCO 3.3.1.1 Condition C, a half scram was generated on RPS trip system 'A' to restore full scram capability. The cause of the failure of the two level instruments associated with RPS Trip system 'A' is under investigation. The level channels are being calibrated prior to changing to mode 1 (power operations). The licensee will notify the NRC Resident Inspector.Reactor Protection System
Reactor Pressure Vessel
ENS 5109428 May 2015 05:17:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessArea Radiation Monitors Non-Functional During Planned Bus OutageA planned outage of the Division 2 medium voltage switchgear (SM-8) was initiated at 22:17 PDT on 5/27/15. The bus outage results in all area radiation monitors required for emergency classification being non-functional. Compensatory measure monitoring equipment has been established prior to the loss to provide alternate means of monitoring area radiation levels. The SM-8 outage window is scheduled to last 124 hours. Although the monitoring function is maintained by the compensatory monitoring equipment, the planned loss of area radiation monitors for greater than 72 hours is being reported as a major loss of emergency assessment capability in accordance with 10 CFR 50. 72(b )(3)(xiii). The NRC Resident Inspector has been notified.
ENS 5108622 May 2015 07:14:0010 CFR 50.72(b)(3)(iv)(A), System ActuationEmergency Diesel Generator Start Due to Actuation of Undervoltage CircuitryAt 0014 PDT on 05/22/2015, Columbia experienced an unexpected momentary loss of SM-7, a Division 1 4.16 kV vital bus, resulting in a start of Emergency Diesel DG-1 . Additionally, under voltage circuitry prevented Standby Service Water pump 1A from starting to support DG-1 in response to the valid under voltage condition, and operators tripped the diesel at 0016 PDT. The SM-7 bus was reenergized by a 115 kV offsite source through backup transformer TR-B. The cause of this event was an inadvertent trip of under voltage circuitry while connecting test equipment in preparation for Diesel and Loss of Power logic testing. Division 1 was inoperable due to ongoing maintenance during the current refueling outage and was not being relied upon for decay heat removal or core circulation. Columbia is in Mode 5 with a coolant temperature of 96 degrees F, water level is at the normal refueling flooded level with fuel pool cooling gates removed. Division 2 is providing required electrical power and supporting components required for decay heat removal and inventory control. There was no impact to Shutdown Safety Assessment. The NRC Resident Inspector has been notified.Emergency Diesel Generator
Service water
05000397/LER-2015-004
ENS 5102730 April 2015 01:11:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessRod Position Indicator System (Rpis) Unplanned OutageAt 1811 PDT on 04/29/2015, the station declared the RPIS system inoperable when a Control Room panel alarmed the loss of indication. The cause of the equipment loss is under investigation. This unplanned equipment outage is being conservatively reported as a major loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). No other safety equipment has been impacted by this event and the plant continues normal operation. The NRC Resident Inspector has been notified.
ENS 510683 April 2015 07:00:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessLoss of Radiological Assessment Capability Due to Non-Functional Radiation MonitorOn 4/21/2015, during performance of source check surveillance on the liquid effluent radiation monitor for the Plant Service Water (TSW), a non-radioactive system, it was discovered that the instrument was determined to be nonfunctional. It was determined on 4/25/15 that the failure was due to an incorrect 'as left' setting from testing conducted on 4/3/2015. The instrument was determined to be non-functional from the period 4/03/15 to 4/25/15 when the setting was corrected. On 5/12/15 it was recognized that because no compensatory measures were implemented during the time the instrument was non-functional that this condition constituted a major loss of radiation assessment capability which is reportable in accordance with 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector will be notified.Service water
ENS 5056923 October 2014 17:46:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessTurbine Building Exhaust Sample Rack Declared Non-FunctionalA planned outage for the Turbine Building Exhaust Air Radiation Indicating Switch (TEA-RIS-13) and the Turbine Building Process Radiation Monitoring Sample Rack (TEA-SR-26) for health inspection was initiated at 1046 PDT on 10/23/14. Due to maintenance retests taking longer than expected and in anticipation of possibly exceeding 72 hours for the planned outage, this event is being reported as a major loss of assessment capability under regulation 10 CFR 50.72(b)(3)(xiii). Compensatory measures have been implemented to obtain radiation readings from the associated effluent release pathway during the outage. Field team assessment function was unaffected and remains available. The Resident Inspector will be notified.
ENS 5038019 August 2014 14:33:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessLoss of Assessment CapabilityThis notification is being made due to a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). At 0733 (PDT), on 08/19/2014, PRM-RE-18, Reactor Building Stack Monitor - Intermediate Range Detector, failed downscale. PRM-RE-1A and PRM-RE-1C, the Reactor Building Stack Monitor - low and high range detectors, both remain operable and fully functional. Compensatory measures are being implemented per plant procedures at this time. The NRC Resident Inspector has been notified.
ENS 503415 August 2014 00:01:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessRadwaste Building Radiation Monitoring Sample Rack Declared Nonfunctional

At 1701 hours PDT on August 4, 2014, the Rad Waste Building process radiation monitoring sample rack was declared nonfunctional. The cause of the equipment malfunction is under investigation. Field team assessment function is unaffected and remains available if required. This event is being reported as a major loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified.

  • * * UPDATE FROM NICHOLAS RULLMAN TO HOWIE CROUCH AT 1501 EDT ON 8/8/14 * * *

At 1501 PDT on 8/7/14, the Rad Waste Building process radiation monitor sample rack was declared functional. The NRC Resident Inspector has been notified. Notified R4DO (Werner).

ENS 5029422 July 2014 13:35:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessRadiation Monitoring Sample Rack Declared Non-Functional

At 0635 hours PDT on July 22, 2014, Turbine Building Exhaust Air Radiation Indicating Switch (TEA-RIS-13) and the Turbine Building Process Radiation Monitoring Sample Rack (TEA-SR-26) were declared non-functional. The cause of the malfunction is under investigation. Compensatory measures have been implemented to obtain radiation readings from the associated effluent release pathway. Field team assessment function was unaffected and remains available. This event is being reported as a major loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM DUANE SALSBURY TO DONALD NORWOOD AT 2018 EDT ON 7/28/2014 * * *

At 1351 PDT on 7/28/2014, Turbine Building Exhaust Air Radiation Indicating Switch (TES-RIS-13) and Turbine Building Process Radiation Monitoring Sample Rack (TES-SR-26) were declared functional. The NRC Resident Inspector has been notified. Notified R4DO (O'Keefe).

ENS 5029221 July 2014 17:03:0010 CFR 26.719, FFD Reporting requirementsFitness for Duty ReportA non-licensed employee supervisor had a confirmed positive for a controlled substance during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee informed the NRC Resident Inspector.
ENS 502594 July 2014 09:33:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessTurbine Building Process Radiation Monitoring Non-FunctionalAt 0233 PDT on July 4, 2014, TEA-RIS-13 and the Turbine Building process radiation monitoring sample rack were declared non-functional. The cause of the malfunction is under investigation. Compensatory measures have been implemented to obtain radiation readings from the associated effluent release pathway. Field team assessment function was unaffected and remains available. This event is being reported as a major loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). The licensee has notified the NRC Resident Inspector.
ENS 501829 June 2014 14:15:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessRadiation Monitoring Sample Rack Declared Non-Functional

At 0715 hours PDT on June 9, 2014, the Rad Waste Building process radiation monitoring sample rack was declared non-functional. The cause of the equipment malfunction is under investigation. Compensatory measures were implemented to obtain radiation readings from the associated effluent release pathway. Field team assessment function was unaffected and remains available. This event is being reported as a major loss of assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). The licensee will notify the NRC Resident Inspector.

  • * * UPDATE PROVIDED BY QUOC VO TO JEFF ROTTON AT 2019 EDT ON 06/14/2014 * * *

At 1255 PDT on 06/13/2014, the Rad Waste Building process radiation monitoring sample rack was declared functional. The NRC Resident Inspector has been notified. Notified R4DO (Werner).

ENS 500843 May 2014 02:20:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessReactor Building Stack Radiation Monitor Failure

This notification is being made due to a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). At 1920 (PDT), on 05/02/2014, PRM-RE-18, Reactor Building Stack Monitor - Intermediate Range Detector, failed downscale. PRM-RE-1A and PRM-RE-1C, the Reactor Building Stack Monitor - Low and High Range Detectors, both remain operable and fully functional. Compensatory measures are being implemented per plant procedures at this time. The NRC Resident Inspector has been notified.

  • * * UPDATE AT 1754 EDT ON 05/08/14 FROM JASON LOVEGREN TO DONG PARK * * *

Following completion of maintenance activities PRM-RE-1B, Reactor Building Stack Monitor - Intermediate Range Detector was returned to operable status at 0810 PDT on 05/08/2014, restoring its required emergency assessment capability. The NRC Resident Inspector has been notified. Notified R4DO (Whitten).

ENS 4989811 March 2014 17:01:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionPostulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown EquipmentAn extent of condition review of all unfused ammeters circuits in Direct Current (DC) distribution systems at Columbia Generating Station (Columbia) identified areas in the plant which may be susceptible to secondary fires due to hot shorts from these unfused ammeters. It is postulated that a fire in one fire area can damage these circuits and cause short circuits without protection that would overheat the cables and possibly result in secondary fires in other fire areas where the cables are routed. The secondary fires could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. This condition is reportable as an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector.05000397/LER-2014-002
ENS 4984119 February 2014 16:29:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessLoss of Primary Containment Oxygen and Hydrogen Monitoring

This notification is being made due to a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). At 0829 (PST), on 2/19/2014 the Division 1 sample rack for monitoring primary containment oxygen and hydrogen atmospheric concentrations was removed from service for planned maintenance activities. The Division 1 sample rack is expected to be out of service for 14 hours. The redundant Division 2 sample rack was previously removed from service for maintenance and remains out of service for repairs. Compensatory measures to monitor primary containment for hydrogen and oxygen are available via grab samples using chemistry procedures. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM NICHOLAS RULLMAN TO VINCE KLCO AT 1400 EST ON 2/20/2014 * * *

Following completion of surveillance activities, the Division 1 sample rack for monitoring primary containment oxygen and hydrogen atmospheric concentrations was returned to operable status at 1037 PST on 2/20/2014, restoring its required emergency assessment capability. The NRC Resident Inspector has been notified. Notified the R4DO (Allen).

Primary containment
ENS 4983417 February 2014 08:44:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Loss of Secondary Containment Differential PressureReactor Building (Secondary Containment) pressure rose above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge multiple times from 0044 (PST) to 0305 (PST) on 2/17/14. The alarm was received in the control room at 0305 (PST). This is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. Reactor Building pressure has been restored to normal (greater than 0.25 inches of vacuum water gauge), returning Secondary Containment to operable status. Highest actual value indicated was +0.21 inches pressure water gauge. The cause of the event is under investigation. There were no radiological releases associated with the event. The NRC Resident Inspector has been notified.Secondary containment
ENS 4976527 January 2014 08:45:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessTurbine and Rad Waste Buildings Radiation Monitors Inoperable Due to Planned Maintenance

At 0045 PST on 1/27/14, the Turbine Building Exhaust Low and Intermediate radiation monitors and the Rad Waste Building Exhaust Low and Intermediate radiation monitors were declared inoperable because of planned maintenance for replacement of the monitors. Compensatory measures have been implemented per station procedures; however, the time required for installation, testing, and acceptance of the new equipment is expected to last several weeks. Therefore, this radiological monitoring equipment outage is being reported as a major loss of assessment capability under regulation 10 CFR 50.72(b)(3)(xiii). Compensatory measures will be in place throughout the duration of the planned equipment outage. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE ON 6/6/2014 AT 2008 EDT FROM JOHN KAINEG TO DONG PARK * * *

At 1416 PDT on 5/31/14, the Turbine Building Exhaust Low and Intermediate radiation monitors were declared operable, and at 1312 PDT on 6/06/14 the Rad Waste Building Exhaust Low and Intermediate radiation monitors were declared operable after outages due to equipment replacements. The NRC Resident Inspector has been notified. Notified R4DO (Taylor).

ENS 4972915 January 2014 17:07:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Loss of Secondary Containment Differential PressureReactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge briefly (5 minutes or less). This is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. Reactor Building pressure has been restored to normal (greater than 0.25 inches vacuum water gauge), returning Secondary Containment to operable status. Highest actual value was 0.21 inches vacuum water gauge. There were no radiological releases associated with the event. The differential pressure change is believed to have been caused by a momentary shift in Heating and Ventilation Systems dampers. The licensee has notified the NRC Resident Inspector.Secondary containment
ENS 4970910 January 2014 01:43:0010 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Loss of Secondary Containment Differential PressureReactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge briefly (two minutes or less) on two occasions. This is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. Reactor Building pressure has been restored to normal (greater than 0.25 inches vacuum water gauge) returning Secondary Containment to operable status. The cause of the event is under investigation. There were no radiological releases associated with the event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector.Secondary containment
ENS 497028 January 2014 18:10:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessLoss of Assessment Capability - Non-Functional Area Radiation Monitors

At (1010 PST) on 1/08/14, during performance of a surveillance the power supply for ten area radiation monitors in the Reactor Building was found with voltage out of specification. As a result, the affected area radiation monitors were declared non-functional. This condition represents a major loss of assessment capability and is being reported as such under 10 CFR 50.72 (b)(3)(xiii). As directed by station procedures, compensatory measures have been enacted until the power supply is restored. The Resident Inspector has been notified.

  • * * UPDATE FROM JASON LOVEGREN TO JIM DRAKE ON 01/10/2014 AT 0214 EST* * *

The power supply voltage has been restored to specification per applicable station procedures. All affected area radiation monitors have been declared functional. Compensatory measures have been suspended. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION ON 3/13/14 AT 1853 EDT FROM JOHN KAINEG TO DONG PARK * * *

Licensee is retracting this event notification based on the following: Energy Northwest performed an evaluation for the reported out-of-specification voltage condition for the power supply to several radiation monitors in the Reactor Building. The evaluation concluded that the voltage deviation from the -24 VDC set point was small and within the calculated uncertainty for the instrument, and did not result in equipment failure. Therefore, it was concluded that the radiation monitors were functional and that the reported major loss of assessment capability did not occur. The licensee has notified the NRC Resident Inspector. Notified R4DO (Farnholtz).

ENS 497007 January 2014 20:10:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessReactor Building Stack Monitor Temporarily Out of Service for Maintenance

At 1210 PST on January 7, 2014 the Reactor Building Stack Radiation Monitor- Intermediate Range detector was declared non-functional due to scheduled maintenance on supporting equipment. The monitor is expected to be out of service for approximately 1 hour. Preplanned compensatory actions have been implemented. This event is being reported as a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). At 1238 PST on January 7, 2014 the Reactor Building Stack Radiation Monitor -Intermediate Range detector was declared functional following scheduled maintenance on supporting equipment. Emergency Assessment Capability has been restored. Preplanned compensatory actions have been secured. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM MATT REID TO DANIEL MILLS AT 1636 EST ON 02/09/2015 * * *

Licensee is retracting this event notification based on the following: Regulatory guidance in NUREG-1022 Revision 3 allows for not reporting EP equipment outages that are planned (i.e., maintenance) when outage time is not expected to exceed or does not exceed 72 hours, and when there are viable compensatory measures in place. Verification of the Control Room Logs indicates Columbia had viable compensatory measures in place during the maintenance outage and the outage duration was less than 72 hours. Columbia met the conditions in NUREG-1022; therefore, this event did not represent a loss of emergency assessment capability. The licensee has notified the NRC Resident Inspector.

Notified R4DO (Allen).

ENS 4966119 December 2013 16:38:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessBoth Divisions of Primary Containment Oxygen and Hydrogen Atmospheric Monitors Inoperable for Maintenance

This notification is being made due to a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). At 0838 PST on 12/19/2013, the Division 1 sample rack for monitoring primary containment oxygen and hydrogen atmospheric concentrations was removed from service for planned maintenance activities. The Division 1 sample rack is expected to be out of service for 14 hours. The redundant Division 2 sample rack was previously removed from service for maintenance and remains out of service for repairs. Compensatory measures to monitor primary containment for hydrogen and oxygen are available via grab samples using chemistry procedures. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM QUOC VO TO VINCE KLCO AT 0130 EST ON 12/20/13 * * *

Following completion of surveillance activities, the Division 1 sample rack for monitoring primary containment oxygen and hydrogen atmospheric concentrations was returned to operable status at 2114 PST on 12/19/2013, restoring its required emergency assessment capability. The NRC Resident Inspector has been notified. Notified the R4DO (Lantz).

Primary containment
ENS 4964517 December 2013 19:33:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessReactor Building Stack Radiation Monitor Inoperable Due to Scheduled Maintenance on Supporting Equipment

At 1133 PST on December 17, 2013, the Reactor Building Stack Radiation Monitor - Intermediate Range detector was declared non-functional due to scheduled maintenance on supporting equipment. Preplanned compensatory actions have been initiated. This event is being reported as a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). A follow up notification will be made when the equipment has been returned to service. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM RUSSELL LONG TO DONG PARK AT 1912 EST ON 12/21/13 * * *

At 1457 (PST) on December 21, 2013, the Reactor Building Stack Radiation Monitor - Intermediate Range detector was declared functional following scheduled maintenance on supporting equipment. Emergency assessment capability has been restored. Preplanned compensatory actions have been secured. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Werner).

ENS 4962612 December 2013 12:36:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessBoth Trains of Primary Containment Hydrogen and Oxygen Monitors Out of Service for Maintenance

This notification is being made due to a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). At 0436 (PST), on 12/12/2013 the Division 1 sample rack for monitoring primary containment oxygen and hydrogen atmospheric concentrations was removed from service for planned maintenance activities. The Division 1 sample rack is expected to be out of service for 14 hours. The redundant Division 2 sample rack was previously removed from service for maintenance and remains out of service for repairs. Compensatory measures to monitor primary containment for hydrogen and oxygen are available via grab samples using chemistry procedures. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM QUOC VO TO HOWIE CROUCH AT 1705 EST ON 12/13/13 * * *

Following completion of maintenance activities, the Division 1 sample rack for monitoring primary containment oxygen and hydrogen atmospheric concentrations was returned to operable status at 1150 PST on 12/13/2013, restoring its required emergency assessment capability. The NRC Resident Inspector has been notified. Notified R4DO (Walker).

Primary containment
ENS 496033 December 2013 16:22:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessReactor Building Stack Radiation Monitor Non-Functional Due to Scheduled Maintenance

During scheduled maintenance, at approximately 0822 PDT on December 3, 2013 the Reactor Building Stack Radiation Monitor - Intermediate Range detector was declared non-functional due to scheduled maintenance on supporting equipment. To compensate for the loss of assessment capability due to the non-functioning radiation monitoring equipment, an additional Health Physics (HP) Technician trained to acquire offsite dose assessment information on offsite releases is available. The additional personnel are pre-staged in support of the radiation monitoring system outage and will be mobilized in accordance with guidance in the compensatory measure instructions. This event is being reported as a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). A follow up notification will be made when the equipment has been returned to service. The licensee has notified the NRC Resident Inspector. The planned outage is scheduled for approximately 90 hours.

  • * *UPDATE PROVIDED BY DAVID HOLICK TO JEFF ROTTON AT 0652 EST ON 12/07/2013 * * *

At 2104 PST on 12/6/13, planned maintenance on the Reactor Building Stack Radiation Monitor - Intermediate Range detector was complete, the instrument was retested satisfactorily, and the instrument was declared functional. This restored the emergency assessment capability in accordance with 10CFR50.72(b)(3)(xiii). The NRC Resident Inspector has been notified. Notified the R4DO(Vasquez) and R1DO(Cook).

ENS 4963125 November 2013 23:00:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentBreach Sizes Exceeded for Control Room Envelope

On December 13, 2013 it was determined that a reportable condition has existed at Columbia Generating Station since 1500 hours (PST) on November 25, 2013. At 1500 hours on November 25, 2013, the Control Room Envelope (CRE) was declared inoperable based on the inability to ensure that the Control Room Emergency Filtration (CREF) System would be able to maintain a positive differential pressure with all areas surrounding the CRE boundary. Columbia does not have the installed instrumentation to directly monitor the differential pressure between the Main Control Room (MCR) and certain areas adjacent to the MCR. The pressure in the adjacent areas is controlled by placing conservative limits on allowed breach size for these adjacent areas. On November 25, 2013, it was identified that the combined breach size associated with several doors in these adjacent spaces resulted in exceeding the allowed limit. Based on exceeding the allowed breach size limit to the adjacent areas, the Control Room Envelope was declared inoperable, and Technical Specification Action Statement 3.7.3.B.1 was entered. An additional breach was discovered on 12/05/13 from a hole in ductwork passing through the cable spreading room, which is one of the adjacent areas to the CRE boundary, and that condition was added to the existing action statement 3.7.3.B.1. These are conditions that could have prevented fulfillment of a safety function of structures that are needed to mitigate the consequences of an accident and are reportable under 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM BRIAN HEAVILIN TO VINCE KLCO ON 1/21/2014 AT 1720 EST* * *

In the case of this event, the Control Room Envelope (CRE) was declared inoperable and Technical Specification Action Statement 3.7.3.B.1 was entered conservatively based upon limited knowledge at the time of discovery. There is no installed differential pressure indication between the CRE and this area adjacent to the Main Control Room (MCR), therefore conservatively this adjacent area is included in the CRE, and leakage is administratively controlled. The leakage between the MCR and the adjacent area included in the CRE exceeded this administrative limit. Testing performed on November 20, 2013, prior to the December 13, 2013 reported events, as well as testing after the event, January 10, 2014, has demonstrated that the leakage identified does not prevent the Control Room Envelope (CRE) from establishing and maintaining the required differential pressure to ensure fulfillment of its required safety function for Control Room Habitability. The ability of the Control Room Emergency Filtration (CREF) system to perform its function of pressurizing and maintaining the Main Control Room positively pressurized with respect to its surroundings was not lost due to the leaking doors and duct specified in the event. Performance of the surveillance without these breaches sealed validated this conclusion. The event described above should not have been reported, as the Control Room Envelope was always operable and capable of fulfilling its safety function with the existing breaches and did not constitute a reportable event as conditions that could have prevented fulfillment of a safety function of structures that are needed to mitigate the consequences of an accident under 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified the R4DO (Spitzberg).