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 Entered dateSiteRegionReactor typeEvent description
ENS 5013025 February 2020 10:40:00Nine Mile PointNRC Region 1GE-5At 0210 (EDT) on May 22, 2014, Nine Mile Point Unit 2, the reactor building vent radiation monitor (Vent WRGMS) was removed from service due to a problem with the check source. The unplanned isolation of Vent WRGMS is a 8-hour report for 10 CFR 50.72(b)(3)(xiii), any event that results in a major loss of emergency assessment capability. Until the equipment is restored, Chemistry will perform sampling requirements per the ODCM. The NRC Resident Inspector has been notified. The licensee notified the State of New York Public Service Commission.
ENS 4954525 February 2020 08:25:00ColumbiaNRC Region 4GE-5

This notification is being made due to a loss of emergency assessment capability. At 0955 (PST) on 11/17/2013,

it was discovered that the CMS-SR-13 sample valves would not align to allow monitoring the drywell. The redundant sample rack, CMS-SR-14, is already out of service due to an issue with the sample pump, which is currently awaiting replacement parts. The capability to monitor the drywell for H2 and O2 remotely is non-existent if an emergency situation were to arise because of these two issues. The only method currently available to sample the drywall for H2 and O2 is manually via chemistry procedures. This method would be unavailable in accident conditions because the sample point isolation valves isolate on an accident signal (F or A). Even if the sample points were not isolated, the Chemist would be unable to draw a sample safely during an accident due to radiation levels in the Reactor Building and/or H2 and O2 levels in containment. PPM 13.1.1 Table 6 specifies DW/WW (Drywell/Wetwell) hydrogen and oxygen measurements as a method for determining potential loss of Primary Containment. PPM 13.14.11 does not specify an alternative method for the loss of CMS-SR-13 & 14. Based on this information and discussions with Licensing and Emergency Preparedness, the station has determined this issue is a loss of emergency assessment capability per 10 CFR 50.72(b)(3)(xiii).

The NRC Resident Inspector has been notified.

  • * * UPDATE FROM DAVID HOLICK TO DONALD NORWOOD AT 0159 EST ON 11/19/13 * * *

At 1351 PST on 11/18/13, repairs on CMS-SR-13 were completed, returning the monitor to operable status of being able to sample both the drywell and wetwell for hydrogen and oxygen as required for emergency assessment capability. The NRC Resident Inspector has been notified. Notified R4DO (PROULX).

ENS 4908425 February 2020 08:12:00Nine Mile PointNRC Region 1GE-5This 60-day telephone notification is being made per the reporting requirements specified in 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to describe an invalid actuation signal affecting containment isolation valves in more than one system. On April 2, 2013, Nine Mile Point 2 (NMP2) received a Division 2 reactor building area high ambient temperature isolation signal when lifting a lead for trip unit E31-N638B while performing surveillance N2-IPS-LDS-Q010, Reactor Building General Area Temperature Instrumentation Channel Functional Test. The isolation signal provided a closure signal to two Reactor Core Isolation Cooling System (RCIC) valves, and three Residual Heat Removal (RHR) system containment isolation valves. As a result of the isolation signal one of the RCIC containment isolation valves, 2ICS*MOV128 closed. The other four valves were already in their normal closed position. The RHR system valves are associated with the RHR Shutdown Cooling System and second RCIC isolation valve is used to warmup and place the RCIC system in standby following an isolated condition. All affected isolation valves responded as designed. As a result of 2ICS*MOV128 closing the RCIC system was declared inoperable. Technical Specification 3.5.3, RCIC System, Condition A was entered. Action A.1 required verifying the High Pressure Core Spray System (HPCS) was operable immediately. Action A.2 requires restoring RCIC to operable within 14 days. After the instrumentation system was restored to normal, the RCIC system was subsequently restored to available later that day at 1205 (EDT) and operable at 1500 (EDT). The actuation signal was not valid because it resulted from maintenance activities when leads were lifted, and the trip unit had not been bypassed as required by the procedure. There were no isolation logic signals in response to actual plant conditions or parameters. This event was entered into the corrective action system as Condition Report (CR) 2013-002461. There were no actual safety consequences or impact on the health and safety of the public as a result of this event. The licensee notified the NRC Resident Inspector and the State.
ENS 4809725 February 2020 07:44:00Nine Mile PointNRC Region 1GE-5On July 12, 2012, at 0200 EDT, clean steam reboiler 'B' failed, causing a loss of sealing steam. This resulted in degrading condenser vacuum and rising off gas system pressure. The main steam backup supply for sealing steam also failed, and condenser vacuum and off gas system pressure continued to degrade. In response to rising off gas system pressure and lowering condenser vacuum, reactor power was lowered to 85% in accordance with Special Operating Procedures. With off gas system pressure approaching the procedural limit and condenser vacuum degrading rapidly, a manual reactor scram was inserted at 0220 EDT. All control rods fully inserted and all systems functioned as expected on the scram. Plant is currently shutdown and parameters are stable. The unit is currently implementing post scram recovery procedures and plant cooldown is in progress. The cause of the loss of both the primary and backup sources of sealing steam is under investigation. The shutdown electrical lineup is normal and decay heat is being removed via steam bypass valves to the main condenser. The licensee has notified the NRC Resident Inspector and the Public Service Commission.
ENS 5341018 May 2018 13:27:00ColumbiaNRC Region 4GE-5At 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.
ENS 5339510 May 2018 08:59:00Nine Mile PointNRC Region 1GE-5At 0248 (EDT), with the plant shutdown in Mode 4, Nine Mile Point Unit 2 experienced a partial loss of off-site power during relay testing that resulted in an automatic start of the Division 2 Emergency Diesel Generator. All systems responded as expected for the event. The cause is being investigated. The station responded in accordance with appropriate Special Operating Procedures and restored impacted systems. This event is being reported in accordance with 10CFR50.72(b)(3)(iv)(A) At the time of the report, the emergency diesel generators are loaded and supplying plant safety equipment. The licensee has notified the state of New York Emergency Management Agency and the NRC Resident Inspector.
ENS 5327622 March 2018 07:07:00LaSalleNRC Region 3GE-5This notification is being provided in accordance with 10CFR50.72(b)(2)(i), Plant Shutdown required by Technical Specifications, and 10 CFR 50.72(b)(3)(ii)A, Degraded or Unanalyzed Condition. At 0300 CDT on 3/22/18, on LaSalle Unit 1, a through-wall (welded joint) leak was identified on a 3/4 inch vent line off of the bonnet of the 1B33-F067B, 1B Reactor Recirculation Pump Discharge Valve. This condition qualifies as pressure boundary leakage, which requires entry into Technical Specification 3.4.5, Reactor Coolant System Operational Leakage, Required Action C, to be in Mode 3, Hot Shutdown, by 1500 on 3/22/18 and Mode 4, Cold Shutdown, by 1500 on 3/23/18. This leakage is significantly less than 10 gallons per minute and therefore, does not meet the threshold for entry into the Emergency Action Plan. At the time of discovery, Unit 1 was in Mode 1 - Run. Shutdown began at 0500 CDT and the estimated completion to cold shutdown is 2000 CDT. All necessary shutdown equipment is available. There is no impact to Unit 2. NRC Resident Inspector was notified.
ENS 5321917 February 2018 10:29:00LaSalleNRC Region 3GE-5This report is being made in accordance with 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. While troubleshooting an issue with the Unit 1B Diesel Generator Oil Circulating pump, damage of a bus bar was identified at the breaker that supplies the Unit 1B Diesel Generator Auxiliaries. One of the loads fed from this breaker is the Division 3 DC Battery Charger. It has been determined that the degradation of the bus bar may have prevented the Division 3 DC Battery Charger from performing its function which could have prevented the High Pressure Core Spray System (HPCS) from performing its design safety function. Since HPCS is a single train safety system, it has been determined that this failure could potentially affect the safety function of this system, and is reportable as an 8 hour notification. The NRC Resident Inspector has been notified.
ENS 5321315 February 2018 13:39:00LaSalleNRC Region 3GE-5On February 15, 2018, during evaluation of protection for Technical Specifications (TS) equipment from the damaging effects of tornado generated missiles, LaSalle Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. Tornado generated missiles could strike the components supporting the operation of Control Room (VC) and Auxiliary Electric Room (VE) ventilation. This could result in inoperable VC/VE systems, which provide a protected environment for occupants to control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke if a tornado were to occur. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v)(D) as 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 mitigate the consequences of an accident. This condition is being addressed in accordance with NRC enforcement guidance provided in Enforcement Guidance Memorandum (EGM) 15-002, Revision 1, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' and DSS-ISG-2016-01, Revision 1, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion' per Enforcement Guidance Memorandum EGM 15-002, 'Enforcement Discretion for Tornado Generated Missile Protection Noncompliance.' Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification.
ENS 531901 February 2018 11:59:00Nine Mile PointNRC Region 1GE-5

Nine Mile Point unit 2 experienced an unusual event due to a small fire in the turbine building that was immediately extinguished and then reflashed. The fire was declared out at 1119 (EST), 2/1/18. The fire was caused when steam leak repair injection equipment failed and leaked onto hot piping. There was no equipment damage or impact to plant operation. The fire was extinguished by the fire brigade. Offsite assistance was not required. The fire resulted from Furmanite repair of a Moisture Separator Reheater inlet flow control valve. The unusual event will be terminated when sufficient lagging is removed to verify the extent of leaked fluid. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA, DHS NICC, and NNSA (via e-mail).

  • * * UPDATE AT 1240 EST ON 2/1/2018 FROM ANTHONY PETRELLI TO MARK ABRAMOVITZ * * *

The unusual event was terminated at 1211 EST. The licensee notified the NRC Resident Inspector. Notified the R1DO (Janda), NRR EO (Miller), IRD MOC (Grant), DHS SWO, FEMA, DHS NICC, and NNSA (via e-mail).

ENS 5308321 November 2017 14:22:00LaSalleNRC Region 3GE-5On October 6, 2017 at 0910 CDT hours, with Unit 1 in Mode 1 (Power Operation), the 1A Diesel Generator Cooling Water Pump (DGCWP) automatically started. The cause was the misoperation of the 1B/C RHR (Residual Heat Removal) Room Cooler Fan (1VY03C) control switch, which was placed in the start position instead of the intended pull-to-lock position. The start of the 1VY03C fan resulted in the automatic actuation of the 1A DGCWP. This system actuation is reportable in accordance with 10CFR50.73(a)(2)(iv)(A). The invalid actuation was not part of a pre-planned sequence during testing or reactor operation. The 1A DGCWP, an emergency service water system that does not normally run and that serves as an ultimate heat sink, responded satisfactorily. This call is being made in accordance with 10CFR50.73(a)(1), which states that in the case of an invalid actuation reported under 10CFR50.73(a)(2)(iv), other than an actuation of the reactor protection system when the reactor is critical, the licensee may provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written Licensee Event Report. The licensee notified the NRC Resident Inspector.
ENS 529993 October 2017 16:58:00ColumbiaNRC Region 4GE-5On 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.
ENS 5299630 September 2017 09:06:00Nine Mile PointNRC Region 1GE-5At 0134 (EDT) on September 30, 2017, Nine Mile Point Unit 2 entered Tech Spec 3.6.4.1 when secondary containment was declared inoperable due to secondary containment differential pressure being above the Tech Spec Surveillance Requirement of -0.25 inches vacuum water gauge. The Division II Standby Gas Treatment System was started to restore differential pressure at 0135 (EDT) on September 30, 2017 the differential pressure was restored, the secondary containment was declared operable and the Tech Spec3.6.4.1 exited. Secondary containment being inoperable is a 8-hour report for 10 CFR 50.72(b)(3)(v)(C), Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. The cause of this condition is being investigated. The NRC Resident Inspector has been notified. The licensee will also inform the State of New York.
ENS 5296612 September 2017 19:11:00ColumbiaNRC Region 4GE-5On 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.
ENS 5291820 August 2017 22:46:00ColumbiaNRC Region 4GE-5

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).

ENS 5291317 August 2017 17:55:00ColumbiaNRC Region 4GE-5Pursuant 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 528896 August 2017 00:26:00Nine Mile PointNRC Region 1GE-5At 2235 (EDT) Nine Mile Point Unit 2 experienced an automatic scram on high reactor pressure. Turbine stop valve testing was in progress at the time of the scram. All control rods inserted. Pressure control is via the turbine bypass valves. The cause of the scram is being investigated. This is a 4-Hour report for 10CFR50.72(b)(2)(iv)(B) RPS (Reactor Protection System) Actuation. The NRC Resident Inspector has been notified. Reactor water level is being maintained with normal feedwater flow. No safety or relief valves lifted. The plant is in its normal shutdown electrical lineup.
ENS 5286924 July 2017 13:36:00ColumbiaNRC Region 4GE-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).
ENS 5284811 July 2017 17:45:00ColumbiaNRC Region 4GE-5

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).

ENS 5282123 June 2017 01:00:00LaSalleNRC Region 3GE-5

This report is being made pursuant to 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 Unit 1 Low Pressure Core Spray (LPCS) Pump Injection System was declared inoperable at 2043 (CDT) due to a loss of corner room area cooling and loss of motor cooling. The common diesel generator cooling water pump received an auto trip signal while being secured. The LPCS pump remained in standby during the event. This condition prevents LPCS, a single train safety system, from performing its design function. This is a reportable condition as an 8 hour ENS notification. The required action of Technical Specifications (TS) 3.5.1, 'ECCS - Operating,' was entered on June 22, 2017 at 2043 CDT when the condition was identified and the LPCS system was determined to be inoperable. Investigation into the cause of the condition is in progress. The Low Pressure Core Spray (LPCS) Pump Injection System was declared operable, and the TS LCO was exited at 2112 CDT.

  • * * UPDATE FROM RICHARD IMMKE TO DONALD NORWOOD AT 1518 EDT ON 6/23/2017 * * *

Update to previous ENS notification at 0100 EDT on 6/23/17. The last statement was revised to say the Low Pressure Core Spray System remains Inoperable. The Low Pressure Core Spray (LPCS) Injection System remains inoperable. The licensee notified the NRC Resident Inspector. Notified R3DO (Pelke).

ENS 5276117 May 2017 12:29:00LaSalleNRC Region 3GE-5This report is being made pursuant to 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 Low Pressure Core Spray (LPCS) Pump Injection HI Flow alarm was received at 09:08 CDT on May 17, 2017, at which point the minimum flow valve was observed to go closed. The LPCS pump remained in standby during the event. To prevent damage if the pump were to auto start, the control switch for the LPCS pump was placed in pull to lock. This condition prevents LPCS, a single train safety system, from performing its design function. This is a reportable condition as an 8 hour ENS notification. The required action of Technical Specifications (TS) 3.5.1, 'ECCS - Operating,' was entered on May 17, 2017 at 09:08 CDT when the condition was identified and the LPCS system was determined to be inoperable. Investigation into the cause of the condition is in progress. There were no related work activities in progress at the time the condition was identified. The licensee notified the NRC Resident Inspector.
ENS 5261516 March 2017 15:07:00ColumbiaNRC Region 4GE-5Pursuant 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 5256418 February 2017 02:58:00LaSalleNRC Region 3GE-5

This notification is being provided in accordance with 10CFR 50.72(b)(2)(iv)(B). On February 17, 2017 at 2353 CST, Unit 1 Reactor Automatic Scram signal was received due to Turbine Control Valves Fast Closure. The turbine trip was due to receipt of Level 8 Trip due to a failure of the Feedwater Regulating Valve to Full open. Plant is in a stable condition with reactor pressure being maintained by the Turbine Bypass valves. Reactor water level is being controlled with Feedwater thru the Low Flow Feedwater Regulating Valve. Further investigation into the cause of the event is in progress. All control rods fully inserted, and decay heat is being removed via steam to the main condenser using bypass valves. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM BROCK POLLMANN TO HOWIE CROUCH AT 1721 EDT ON 4/14/17 * * *

Upon further review of the event data, it was determined that the Nuclear Station Operator (NSO) had initiated a manual scram, which was followed by a Turbine Control Valve (TCV) fast closure automatic scram when the turbine tripped. The licensee has notified the NRC Resident Inspector. Notified R3DO (Jeffers).

ENS 5254714 February 2017 02:40:00LaSalleNRC Region 3GE-5This notification is being provided in accordance with 10 CFR 50.72(b)(2)(iv)(B). On February 13, 2017 at 2309 CST, a Unit 1 Reactor Automatic Scram signal was received due to Turbine Control Valve Fast Closure. The turbine trip was due to the main generator trip on Differential Current. The 'U' safety relief valve actuated in the relief mode on the turbine trip, and has subsequently reset with tailpipe temperature returning to normal. The plant is in a stable condition with reactor pressure being maintained by the Turbine Bypass valves. Reactor water level is being controlled with feedwater. Investigation into the cause of the event is in progress. All control rods fully inserted on the scram. Unit 1 is in a normal shutdown electric plant alignment. Unit 2 was not affected. The licensee notified the NRC Resident Inspector.
ENS 5251930 January 2017 23:34:00LaSalleNRC Region 3GE-5This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D), event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. During routine surveillance testing of the Unit 2 Division 3 Emergency Diesel Generator (LOS-DG-M3), the Cooling Water Strainer Backwash Valve, 2E22-F319, was identified to have stem/disk separation and could not be opened. This condition has been evaluated and the Division 3 Diesel Generator Cooling Water system has been declared inoperable. The Division 3 Diesel Generator Cooling Water system is a support system for the Division 3 Emergency Diesel Generator and the High Pressure Core Spray System (HPCS). The required actions of Technical Specification (TS) 3.5.1 were entered on 1/30/17 at 1908 CST when the HPCS system was determined to be inoperable. This condition could have prevented the HPCS, a single train safety system, from performing its design function. The licensee has notified the NRC Resident Inspector.
ENS 5251026 January 2017 03:11:00ColumbiaNRC Region 4GE-5On 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.
ENS 524722 January 2017 17:49:00LaSalleNRC Region 3GE-5At approximately 1403 (CST) on January 2, 2017, LaSalle County Station was informed by the LaSalle County Sheriff that two warning sirens had malfunctioned at approximately 1252 earlier this afternoon. The warning sirens had inadvertently operated for nearly 3 minutes during maintenance at the LaSalle County 911 Communications Center. The warning sirens have been restored to standby and are fully functional. No other emergency notification systems or sirens were affected. The licensee has notified the NRC Resident Inspector of the issue.
ENS 5238220 November 2016 18:53:00ColumbiaNRC Region 4GE-5On 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.
ENS 522763 October 2016 16:49:00ColumbiaNRC Region 4GE-5On 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.
ENS 522278 September 2016 14:50:00LaSalleNRC Region 3GE-5On July 11, 2016, at approximately 0430 CDT, while Unit 1 was operating at 100% power, the 1A Reactor Protection System (RPS) Motor Generator (M/G) set tripped causing a loss of the A RPS bus. This caused the complete actuation of the Division 1 (outboard) primary containment isolation logic. The isolation logic actuation resulted in successful closure of the Division 1 primary containment isolation valves. This was an event that resulted in the actuation of a general containment isolation signal affecting more than one system. However, as this event meets the definition of an invalid actuation (i.e., not a response to an actual plant parameter exceeding a trip set-point), this notification is being made in accordance with 10CFR50.73(a)(2)(iv)(A) in lieu of a Licensee Event Report. In response to the trip of the 1A RPS M/G Set, operators swapped the A RPS bus to the alternate power supply using the applicable response procedure. The containment isolation signal was reset and the systems were restored to their normal lineup. Reactor power was not affected by this event. All safety related equipment controlled by the affected primary containment isolation circuits operated as designed. The 1A RPS M/G Set trip was due to a blown power fuse for the 1A RPS M/G Set. This was the result of worn insulation on one of the generator output leads. The generator output leads were repaired and rerouted to prevent future problems on 07/15/16. Restoration of the normal power supply to the 1A RPS function was completed on 7/19/16. The licensee informed the NRC Resident Inspector.
ENS 5210418 July 2016 23:29:00LaSalleNRC Region 3GE-5This telephone notification is provided in accordance with the Exelon Reportability manual, 'Major Loss of Emergency Preparedness Capabilities', and 10 CFR 50.72(b)(3)(xiii). On July 18, 2016 at 1500 CDT, it was determined during testing of the Everbridge ERO (Emergency Response Organization) notification system that the system would not notify the corporate EOF (Emergency Operations Facility) individuals if the system had been activated. This constitutes a loss of offsite communication capability. Exelon and Everbridge have identified and corrected this issue. A follow-up test of the LaSalle Everbridge ERO notification system was completed satisfactorily on July 18, 2016 at 2100. The Emergency Response Data System (ERDS) capability was not lost. The licensee has notified the (NRC) Senior Resident Inspector of the issue. Compensatory measures were in place during this event.
ENS 5209618 July 2016 14:48:00Nine Mile PointNRC Region 1GE-5
GE-2

This 8-hour non-emergency report is being made based upon requirements of 10 CFR 50.72(b)(3)(xiii) which states, 'The licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' On 07-18-2016 at 0730 (EDT), both Control Rooms were notified by the Emergency Preparedness Manager, that the Everbridge Emergency Response Organization (ERO) Notification System may not notify all ERO individuals within the required 10 minutes of system initiation. This constitutes a loss of offsite communications capability. The Everbridge vendor is working to resolve the issue. Compensatory measures are in place. All ERO personnel received the page but not all received the notification within the required ten minutes. The licensee notified the NRC Resident Inspector and the State.

  • * * UPDATE AT 1728 EDT ON 7/20/2016 FROM CLARK WILLETT TO MARK ABRAMOVITZ * * *

The Everbridge System was restored and retested at 1930 EDT on 7-19-2016 to provide offsite communications capability. The licensee notified the NRC Resident Inspector and the state. Notified the R1DO (DeFrancisco).

ENS 518961 May 2016 12:25:00Nine Mile PointNRC Region 1GE-5
GE-2
On May 1, 2016, at 0847 (EDT), an individual experienced a personal medical emergency during a break. The onsite fire brigade and emergency medical technicians administered first aid, but the individual was unresponsive. The individual was transported to the local hospital. The station was notified at 1008 that the hospital has declared the individual deceased. The individual was outside of the radiological controlled area and not contaminated. Nine Mile Point Unit 2 is shut down for the scheduled refueling outage. The individual was a contractor employee. The licensee has notified the NRC Resident Inspector. The State of New York will be notified.
ENS 518567 April 2016 22:32:00Nine Mile PointNRC Region 1GE-5Nine Mile Point Unit 2 (NMP2) experienced a momentary loss of Secondary Containment due to both Reactor Building (RB) airlock doors being opened at the same time. At 1730 (EDT) on 04/07/16, both RB airlock doors were opened simultaneously for less than 5 seconds. This resulted in Secondary Containment being declared Inoperable (TS 3.6.4.1). Secondary Containment was restored to Operable when the doors were closed. The event is reportable in accordance with 10 CFR 50.72(b)(3)(v)(C), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.' The condition has been entered into the station's corrective action program and the NRC Resident Inspector has been notified. There is no interlock on these doors, just lights to verify the opposite door is open or shut. An Operator was entering secondary containment as another Operator was leaving.
ENS 5172511 February 2016 01:20:00LaSalleNRC Region 3GE-5This report is being made pursuant to 10CFR50.72(b)(3)(v)(C), event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and 10CFR50.72(b)(3)(v)(D), event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. LaSalle Station's Unit 1 and Unit 2 were in Mode 1. At 2207 (CST) (on 2/10/16), Secondary Containment Differential Pressure dropped below the Technical Specification (TS) 3.6.4.1 minimum of 0.25 inches water vacuum. The initial indications are a failure of one Unit 1 Reactor Building Exhaust Isolation Damper, which resulted in a trip of the Unit 1 Reactor Building Exhaust Fans. At 2245, Secondary Containment Differential Pressure was restored to within the TS 3.6.4.1 limits by securing and isolating the Unit 1 Reactor Building Ventilation System. Troubleshooting plans are being developed to determine cause of the damper failure and to correct the deficient condition. The licensee has notified the NRC Resident Inspector.
ENS 5172310 February 2016 06:29:00Nine Mile PointNRC Region 1GE-5
GE-2
At approximately 0354 (EST) on 2/10/2016, the Nine Mile Point Unit 1 Control Room was notified by Exelon Emergency Preparedness of the inadvertent actuation of one Oswego County Notification Siren at approximately 0247 on 2/10/2016. It is unknown at this time why the inadvertent alarm occurred. Siren repair personnel (ANS Services) have been dispatched to isolate the siren and begin repair work. The siren has been silenced. Alternate notification of the public in the area is through Hyper Reach. The Oswego County Emergency Management Office has issued a news release identifying the inadvertent actuation of the emergency siren. The licensee has notified the NRC Resident Inspector.
ENS 5156223 November 2015 17:50:00ColumbiaNRC Region 4GE-5At 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 5153011 November 2015 19:40:00LaSalleNRC Region 3GE-5At 1344 CST on 11/11/15, the seismic monitor was found inoperable. The seismic monitor was inoperable such that emergency classification at the ALERT level could not be obtained with site instrumentation. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). (The NRC) Senior Resident Inspector has been notified.
ENS 5152610 November 2015 04:15:00ColumbiaNRC Region 4GE-5At 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.
ENS 513007 August 2015 19:58:00LaSalleNRC Region 3GE-5This notification is being provided in accordance with 10 CFR 50.72(b)(3)(ii)A, Degraded Condition. At 1340 CDT on 8/7/15, on LaSalle Unit 2, a through-wall (welded joint) leak was identified on a 3/4 inch vent line off of the bonnet of the 2B33-F067B, 2B Reactor Recirculation Pump Discharge Valve. This condition qualifies as pressure boundary leakage, which requires entry into Technical Specification 3.4.5, Reactor Coolant System Operational Leakage, Required Action C, to be in Mode 4, Cold Shutdown, by 0140 on 8/9/15. This leakage is significantly less than 10 gpm (leak rate is 0.2 gpm) and therefore, does not meet the threshold for entry into the Emergency Action Plan. At the time of discovery, Unit 2 was in Mode 3 - Hot Shutdown, heading into Cold Shutdown for a planned maintenance outage. This event does not affect Unit 1. The licensee notified the NRC Resident Inspector.
ENS 5124722 July 2015 07:08:00LaSalleNRC Region 3GE-5(At) 0013 CDT, (on) 7/22/15, the seismic monitor was found inoperable. The seismic monitor was inoperable such that emergency classification at the ALERT level could not be obtained with site instrumentation. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). (The NRC) Senior Resident Inspector has been notified.
ENS 512138 July 2015 21:53:00LaSalleNRC Region 3GE-5

This telephone notification is provided in accordance with Exelon Reportability manual SAF 1.10, 'Major Loss of Emergency Preparedness Capabilities', and 10CFR50.72(b)(3)(xiii). On July 8th 2015 at 1837 (CDT), it was determined that the onsite Technical Support Center (TSC) Ventilation System Supply Fan belts had failed, resulting in loss of ventilation for the facility. Repairs were not completed within the time required had the TSC needed to be staffed. There is currently no emergency event in progress requiring TSC staffing. If an emergency is declared and the TSC ERO (Emergency Response Organization) activation is required, the TSC will be staffed and activated unless the TSC becomes uninhabitable due to ambient temperatures, radiological, or other conditions. If relocation of the TSC staff becomes necessary, the Station Emergency Director will relocate the staff to an alternate TSC location in accordance with applicable site procedures. The licensee has notified the (NRC) Senior Resident Inspector of the issue.

  • * * UPDATE FROM TODD CASAGRANDE TO DANIEL MILLS AT 1510 EDT ON 7/11/15 * * *

After repairs were completed, the TSC Ventilation was restarted on 7/9/15 at 0625 EDT for a maintenance run, the TSC Ventilation was restored to operable status at 1500 EDT on 07/11/2015. The licensee has notified the NRC Resident Inspector. Notified R3DO (Stone).

ENS 512058 July 2015 07:04:00LaSalleNRC Region 3GE-5At 0130 (CDT) on 7/8/15, the seismic monitor was found inoperable. The seismic monitor was inoperable such that emergency classification at the Alert level could not be obtained with site instrumentation. The seismic monitor was restored to operable status within 11 minutes. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). The licensee has notified the NRC Resident Inspector.
ENS 512016 July 2015 20:41:00ColumbiaNRC Region 4GE-5A 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.
ENS 5118226 June 2015 04:38:00ColumbiaNRC Region 4GE-5At 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.
ENS 5116417 June 2015 23:39:00LaSalleNRC Region 3GE-5

On June 17th, 2015 at 1841 CDT, it was determined that the onsite Technical Support Center (TSC) Ventilation System return damper 0VS119Y was failed closed, the failed closed damper affects the TSC Emergency Makeup Train filtration efficiency. There is currently no emergency event in progress requiring TSC staffing. If an emergency is declared and the TSC ERO (Emergency Response Organization) activation is required, the TSC will be staffed and activated unless the TSC becomes uninhabitable due to ambient temperatures, radiological, or other conditions. If relocation of the TSC staff becomes necessary, the Station Emergency Director will relocate the staff to an alternate TSC location in accordance with applicable site procedures. The licensee has notified the (NRC) Senior Resident Inspector of the issue.

  • * * UPDATE AT 1700 EDT ON 06/18/15 FROM TODD CASAGRANDE TO S. SANDIN * * *

After repairs were completed, the TSC Ventilation was restored to service at 1650 EDT on 06/18/2015. The licensee has notified the NRC Resident Inspector. Notified R3DO (Pelke).

ENS 511336 June 2015 17:22:00LaSalleNRC Region 3GE-5

The following was received via fax and phone: This telephone notification is provided in accordance with Exelon Reportabllity manual SAF 1.1 0, 'Major Loss of Emergency Preparedness Capabilities', and 10CFR50.72(b)(3)(xiii). On June 6th 2015 at 12:13 (CDT), it was determined that the onsite Technical Support Center (TSC) Ventilation System Supply Fan belts had failed, resulting in loss of ventilation for the facility. Repairs were not completed within the time required had the TSC needed to be staffed. There is currently no emergency event in progress requiring TSC staffing. If an emergency is declared and the TSC ERO activation is required, the TSC will be staffed and activated unless the TSC becomes uninhabitable due to ambient temperatures, radiological, or other conditions. If relocation of the TSC staff becomes necessary, the Station Emergency Director will relocate the staff to an alternate TSC location in accordance with applicable site procedures. The licensee has notified the Senior Resident Inspector of the issue.

      • UPDATE PROVIDED BY TODD CASAGRANDE TO NESTOR MAKRIS AT 2305 EDT ON 06/06/2015 ***

After repairs were completed, the TSC ventilation was restored to service at 2300 (EDT) on 06/06/2015. The licensee has notified the NRC Resident Inspector. Notified R3DO (Passehl) via email

ENS 5109428 May 2015 03:35:00ColumbiaNRC Region 4GE-5A 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 10:43:00ColumbiaNRC Region 4GE-5At 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.
ENS 5106812 May 2015 22:13:00ColumbiaNRC Region 4GE-5On 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.