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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).

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).

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).

During performance of surveillance OSP-CONT-M102, secondary containment had been declared inoperable and Technical Specification 3.6.4.1.A had been entered. Operations personnel performing the surveillance were holding the inner door open and testing the outer door interlock. When the individual presented his badge to the badge reader and attempted to open the outer door, the door opened. Per the surveillance contingency actions, the individual immediately shut the door. Momentarily, both the inner and outer reactor building access doors were open. The surveillance was completed with no further issues. Technical Specification 3.6.4.1.A was exited and secondary containment was declared operable. There were no radiological releases associated with this event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector.

• * * RETRACTION FROM DIEGO SUAREZ TO DONALD NORWOOD AT 17:55 EDT ON 3/25/2013 * * *

The licensee is retracting this event based on the following: During subsequent evaluation for reportability, guidance in NUREG-1022 was reviewed. NUREG-1022 provides that removal of a system or part of a system from service as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that could have prevented the system from performing its function) is generally not reportable. In the case of this event, the secondary containment system was removed from service as part of a planned surveillance test. TS Action 3.6.4.1.A was entered as directed by the governing surveillance procedure in anticipation of the airlock interlock testing which, by the very nature of the test, could result in both doors in a given airlock being open simultaneously.

The event described above should not have been reported in consideration of the guidance provided in NUREG-1022 and did not constitute a reportable event as a condition that could have prevented the fulfillment of a safety function under 10CFR50.72(b)(3)(v). The licensee will notify the NRC Resident Inspector. Notified R4DO (Pick).

At 1650 PDT Columbia Generating Station entered LCO 3.0.3 as required by LCO 3.7.3 for both divisions of Control Room Emergency Filtration (CREF) (being) inoperable. Entry into LCO 3.0.3 was required by Condition E of LCO 3.7.3, 'Two CREF subsystems inoperable in MODE 1, 2, or 3 for reasons other than Condition B' and has a Required Action to enter LCO 3.0.3 with an immediate Completion Time. Columbia commenced a plant shutdown at 1747 as required by LCO 3.0.3. (Columbia Generating Station) entered into Condition E of LCO 3.7.3 when it was determined that Radwaste Building Outside Air Smoke Detectors WOA-SMD-1A and WOA-SMD-1B did not meet seismic qualifications. Actuation of these smoke detectors would cause Radwaste Building Mixed Air Fire Dampers WMA-FD-1 and WMA-FD-2 to close. These fire dampers shutting would cause all CREF airflow to cease, preventing the control room from achieving a positive pressure. Both divisions of CREF (being) inoperable is an unanalyzed condition requiring immediate entry into LCO 3.0.3. This event is reportable under 50.72(b)(2)(i) due to the initiation of a plant shutdown as required by Columbia's Technical Specifications. This event is also reportable under 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. WMA-FD-1 and WMA-FD-2 have been deactivated in the open position for the smoke actuation function. (Columbia Generating Station) entered RFOLCS (Requirements for Operability of Licensee Controlled Specifications) 1.10.6 for these fire dampers (being) nonfunctional. (Columbia Generating Station) completed the required actions of LCS 1.10.6. The CREF system remains inoperable pending completion of a TMR (Temporary Modification Request). The licensee notified the NRC Resident Inspector.

* * * UPDATE FROM QUOC VO TO CHARLES TEAL ON 4/7/10 AT 1957 * * * 

Question posed by NRC at time of initial report: Have these smoke detectors been installed since startup? Response: These smoke detectors have been installed since initial plant startup. A temporary modification request was installed to de-energize the non-seismic detector logic that causes automatic fire damper actuation for the two dampers identified above. The alarm function of the smoke detectors remains active with this temporary modification. Entry into this condition was made when it was determined that the smoke detectors were not seismically qualified and there is a possible failure mode that during an earthquake the detectors would actuate the electro-thermal link and cause the associated fire damper (WOA-FD-1 or WOA-FD-2) to close, hence impacting control room emergency filtration function. With implementation of the temporary modification at 2218 PDT on 4/6/10, the CREF system is now considered to be in an 'operable but non-conforming' status. As such, LCOs 3.7.3 and 3.0.3 were exited at 2218 PDT on 4/6/10. This event was initially reported pursuant to paragraph 50.72(b)(2)(i) as an initiation of a plant shutdown required by Columbia's Technical Specifications. Reporting pursuant to paragraph 50.72(b)(2)(i) is being retracted because negative reactivity was not inserted into the reactor prior to exiting the LCOs described above. Reporting pursuant to paragraph 50.72(b)(2)(i) would not be consistent with the guidance provided in NUREG-1022 in that no negative reactivity was inserted and the condition was corrected before exceeding the required action times. The licensee notified the NRC Resident Inspector.

* * * RETRACTION FROM KYLE CHRISTIANSON TO PETE SNYDER AT 1550 EDT ON 4/29/2010 * * *

Subsequent to the April 7, 2010, event report (EN #45816), an Energy Northwest engineering evaluation concluded that the identified smoke detectors would not inadvertently actuate during a design basis safe shutdown earthquake (SSE). This conclusion is based on a seismic response spectrum test report for the smoke detectors. Therefore, since the safety function of the CREF system needed to mitigate the consequences of an accident would not be affected by an SSE, the event notification made pursuant to 10 CFR 50.72 (b)(3)(iv)(D) is hereby retracted. The NRC Senior Resident Inspector (Cohen) has been notified. Notified R4DO (Clark).

The licensee provided the following information: This event notification is being made to report an event that could have prevented fulfillment of the safety function to mitigate the consequences of an accident IAW 10CFR50.72(b)(3)(v)(D). The potential for creation of an actual hole through Secondary Containment (SC) via the Plant Service Water (TSW) system high point reactor building auto vents (TSW-AV-1A and TSW-AV-1 B) exists if a seismic event occurs and the Seismic Category 2 TSW loop-seal piping outside of Secondary Containment is breached (e.g., because of a pipe break) and has drained. Neither of the above conditions presently exists. However, a Secondary Containment breach could occur as a result of a single passive failure (i.e., pipe break in the TSW loop-seal piping described above). TSW-AV-1A and TSW-AV-1B are designed to open automatically when neither TSW system pump is operating to break the vacuum condition that would otherwise exist in the piping. When TSW-AV-1A and TSW-AV-1B are open in vacuum breaker mode, the resultant hole size into the TSW system piping would exceed the allowable Secondary Containment cumulative hole size ( of 32 square inches total), if the loop seal were also breached as described above. Due to this condition Secondary Containment was declared INOPERABLE at 1800 PST 2/24/05. As a compensatory measure to prevent exceeding the allowable Secondary Containment cumulative hole size, one of the two TSW reactor building auto vents (TSW-AV-1B) was isolated by closure of a manual valve (TSW-V-55B). This action was completed at 1828 PST 2/24/05 and the Secondary Containment was declared OPERABLE. One TSW auto vent is sufficient to perform the vacuum breaker function. This condition was found by the licensee's System Engineer. The extent of condition and long term corrective action is under review by licensee. Licensee will inform the NRC Resident Inspector.

• * * RETRACTION FROM FRED SCHILL TO HOWIE CROUCH @ 1535 EDT ON 04/05/05 * * *

The following information was obtained from the licensee via facsimile (licensee text in quotes): On 2/25/05, Columbia Generating Station reported (ref: EN 41442) a condition that was discovered while reviewing service water (TSW) system design documents. During the review, it was determined that Seismic Category II TSW piping in the turbine (TG) and radwaste (RW) buildings could rupture and drain during a seismic event. This event would result in an inoperable secondary containment (SC) because TSW system high point vent valves located within the SC would automatically open when the piping drained after rupturing. Such an event would allow direct communication between the SC atmosphere and the TG/RW atmospheres and exceed the leakage rate assumed in Columbia's accident analysis. There are two principal accidents in Columbia's safety analysis for which SC is credited as a mitigating system. These are the Loss Of Coolant Accident (LOCA) and the Fuel Handling Accident (FHA). The SC performs no active function in response to either of these limiting events, however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis, and that fission products entrapped within the SC structure will be treated prior to discharge to the environment. Assuming a seismic event of the magnitude (0.25g) of the Safe Shutdown Earthquake (SSE) occurred and ruptured the TSW piping causing SC to become inoperable, it is beyond Columbia's safety analysis to postulate a release of radioactive material beyond Part 100 limits for that event. This is because analysis shows that the SSE will not in itself cause a LOCA or an FHA and Columbia's design and licensing bases do not assume a LOCA or FHA coincident with a seismic event. In the event that the SSE made SC inoperable, the Technical Specifications (LCO 3.6.4.1.B) require that the plant be in mode 3 in 12 hours and in mode 4 in 36 hours. Since this is achievable, it is reasonable to conclude that, in the event previously reported, plant shutdown can be accomplished without radiological release and within the completion time of the action required by the Technical Specifications. The discussion in the guidance document (NUREG 1022) for reporting under Part 50.72(b)(3)(v) states the level of judgment for reporting under these criteria is a reasonable expectation of preventing fulfillment of a safety function. It also states that the intent of the criteria is to capture those events where there would have been a failure of a safety system to properly complete a safety function regardless of whether there was an actual demand. This discussion however, as the Part 50.72(b)(3)(v) and Part 50.73(a)(2)(v) criteria state, apply to safety functions of systems or structures that are needed to control the release of radioactive material, because safe shutdown of the plant without radiological release is assured post SSE (sans SC), as described in Columbia's design and licensing bases, SC would not be needed to control the release of radioactive material and therefore the reporting criteria is not met. The licensee will be voluntarily submitting a Licensee Event Report as a method of information sharing with the rest of the industry. The licensee has notified the NRC Resident Inspector. Headquarters Operations Officer notified R4DO (Howell).

The High Pressure Core Spray (HPCS) system was declared inoperable due to the pump's failure to meet the flow requirement specified in TS Surveillance Requirement 3.5.1.4. This surveillance is normally performed on a quarterly basis in accordance with the plant's In-service Testing (IST) Program. The flow values measured during the performance of this surveillance were below both the normal and alert ranges. This test had also been run on 5/27/04 with results in the alert range; HPCS system instruments had been vented between the two tests to rule out the possibility that the results were due to measurement errors. Upon declaring the HPCS pump inoperable, TS 3.5.1 Action B was entered. In accordance with Action B, the Reactor Core Isolation Cooling (RCIC) system was verified to be operable. With the RCIC system verified operable, Action B provides a 14-day completion time to restore HPCS to an operable status. All other Emergency Core Cooling Systems (ECCS) were operable during this event. This event is being reported as an event or condition that could have prevented the fulfillment of a safety function credited for mitigating the consequences of an accident. The HPCS system is a single train system at Columbia. The licensee notified the NRC Resident Inspector.

• * * UPDATE ON 06/14/04 @1752 BY MIKE BRANDON TO C. GOULD * * * RETRACTION

On May 30, 2004, Energy Northwest provided an 8-hour notification pursuant to 10 CFR50.72(b)(3)(v)(D). This notification reported the apparent failure of the High Pressure Core Spray (HPCS) system's pump to meet the flow requirements of Surveillance Requirement (SR) 3.5.1.4. Upon the apparent failure to satisfy this SR, Energy Northwest entered Action B of Technical Specification (TS) 3.5.1 (14 day completion time) and initiated actions to investigate the cause of this apparent failure. This investigation determined the cause of this apparent failure was due to an anomaly in the processing of the pressure and flow input signals and the instrumentation used for documenting the results of the surveillance. Additional testing using alternative instrumentation determined the HPCS pump was fully capable of providing flow within the existing acceptance criteria of the plant's In-service Testing (IST) Program and thus capable of satisfying the SR. This investigation determined that no actual degradation of the pump existed that would have caused a valid failure of the SR. This was an instrumentation issue only. The HPCS system would have been capable of performing its specified safety function in the as-found condition and was capable of fully satisfying the SR. Therefore, this condition would not have prevented the fulfillment of a safety function and is therefore not reportable under 10CFR50.72. The HPCS system was declared OPERABLE on June 03, 2004 at 22:59 PDT. The NRC Resident Inspector was notified.