Semantic search
| Discovered date | Reporting criterion | Title | Event description | |
|---|---|---|---|---|
| ENS 49152 | 28 June 2013 00:58:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray and Division 3 Diesel Generator Inoperable for 20 Minutes | At 1758 (PDT) on June 27, 2013, an alarm signaling heating, ventilation and air conditioning (HVAC) trouble in the Division 3 diesel and high pressure core spray (HPCS) room was received in the main control room. Follow-up investigation determined that the switch for the normal room supply fan (DMA-FN-32) was off. It is believed that the switch may have been inadvertently mispositioned during ongoing work in the vicinity. A worker was moving a vacuum nearby and stopped when he heard the local alarm. Columbia Generating Station is performing further investigation to determine if there are other possibilities for DMA-FN-32 control switch being mispositioned. The fan was returned to service at 1819 (PDT). An Operation's supervisor was present when returning the switch to ON and verified the switch operated as expected. Loss of DMA-FN-32 results in both the HPCS diesel (DG-3) and the HPCS system being inoperable due to inadequate cooling of those systems. Offsite power for Division 3 was verified to be operable while DG-3 was inoperable. The loss of the HPCS system results in the loss of safety function for a single train system and thus is reportable under the 10 CFR 50.72 sections noted above. Appropriate Technical Specification actions were entered and exited for DG-3 and HPCS inoperability times. There was no radiological release associated with this event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector. |
| ENS 49167 | 1 July 2013 22:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Minimum Flow Valve Pressure Switch Setpoint Found Outside of Tolerance | This report is being made pursuant to 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. During routine Instrument Maintenance Surveillance Testing (LIS-HP-205), the High Pressure Core Spray System (HPCS) minimum flow valve pressure switch set point was found outside the Technical Specification allowable value. This could have prevented the High Pressure Core Spray System (HPCS), a single train safety system, from performing its design function. This is reportable as an 8 hour ENS notification. The required actions of Technical Specification (TS) 3.5.1 were entered on 7/01/13 at 1646 (CDT) when the system was made inoperable for surveillance testing. At 1705, maintenance personnel reported minimum flow valve pressure switch set point was found at 112.6 psig, which is outside of the TS Allowable Value of greater than or equal to 113.2 psig (0.6 psig below the Allowable Value). The minimum flow valve pressure switch set point has been calibrated and was left within Technical Specification allowable values, HPCS was declared OPERABLE at 1815 on 7/01/13." The licensee has notified the NRC Resident Inspector.
The event notification was reported by LaSalle Generating Station on 7/01/2013 at 2147 EDT. This update is being provided for the purposes of retracting that notification. On July 1, 2013, during surveillance testing, the Unit 2 High Pressure Core Spray System (HPCS) minimum flow valve pressure switch setpoint was found below the Technical Specification (TS) 3.3.5.1 allowable value. HPCS was declared inoperable, and TS Required Actions (RA) were entered on July 1, 2013, at 1646 hours (CDT). Because HPCS is a single train system, an ENS report was 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 needed to mitigate the consequences of an accident. The minimum flow valve pressure switch setpoint was calibrated to within TS allowable values, and HPCS was declared operable on July 1, 2013, at 1815 hours. A post-event review determined that declaring HPCS inoperable was not required. The inoperability of the pressure switch would not have impacted the function of the HPCS minimum flow valve to automatically open as required to prevent overheating of the HPCS pump. The as-found setpoint was 0.6 psig below the TS allowable value, which would have resulted in the minimum flow valve opening slightly sooner. The inoperability would also not have prevented or delayed the automatic closing of the valve at the required system flow to assure that adequate ECCS flow is available. It should be noted that the LaSalle ECCS (Emergency Core Cooling System) LOCA (Loss Of Coolant Accident) analysis assumes that the HPCS minimum flow valve is open during an injection. TS 3.3.5.1 RA D.4 requires that the minimum flow valve pressure switch be restored to operable status within 7 days. If it cannot be restored within that time, RA G.1 requires that the supported system (HPCS) be declared inoperable, precluding extended operation with the minimum flow pressure switch inoperable. The pressure switch was re-calibrated to within TS allowable values within approximately 1 hour and 29 minutes of being declared inoperable. Therefore, the HPCS system was operable with the minimum flow pressure switch 0.6 psig out of calibration for 1 hour and 29 minutes. This event did not constitute a loss of safety function of the HPCS system, and the event was not reportable under 10 CFR 50.72(b)(3)(v)(D). The licensee has notified the NRC Resident Inspector. Notified R3DO (Lara). |
| ENS 49286 | 15 August 2013 14:15:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Diesel Generator Declared Inoperable Due to Damper Failing to Open | During a run of the Division 3 Diesel Generator room ventilation fan to perform thermograph, it was identified that the damper (1VD01YC) that provides the flow path from the outside area into the ventilation room would not open when the fan was started. This renders the Division 3 Diesel Generator inoperable. High Pressure Core Spray was declared inoperable at 1420 hours (CDT), but remains available. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) as an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. The cause of the damper failing to open has not yet been determined. Troubleshooting is in progress to determine the cause and actions required to restore operability. The Division 1 and Division 2 Diesel Generators are operable. The licensee is in a 14-day shutdown TS LCO. The licensee has notified the NRC Resident Inspector. |
| ENS 49859 | 27 February 2014 08:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray (Hpcs) System Declared Inoperable | On February 24th at 0400 (EST), the Division 3 diesel (HPCS diesel) was declared inoperable for planned maintenance. Technical Specification (TS) 3.8.1 condition B was entered with a required action to restore the diesel to operable status within 72 hours. Shortly after starting the diesel for post maintenance testing, operations observed erratic voltage regulator operation. The diesel was secured at 1621 on 2/24/14 before completing the required post maintenance operability test and troubleshooting is ongoing. At 0330 on 2/27/14, the High Pressure Core Spray System (HPCS) was declared inoperable and TS 3.5.1 condition B was entered. With the HPCS system declared inoperable, TS 3.8.1 condition B was exited in accordance with the following note modifying TS 3.8.1: 'Division 3 AC electrical power sources are not required to be OPERABLE when High Pressure Core Spray (HPCS) System is inoperable.' The HPCS system is a single train system that is discussed in Chapters 6 and 15 of the Final Safety Analysis Report. The unplanned inoperability of the HPCS system is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D) as, '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: (D) Mitigate the consequences of an accident.' The condition has been entered into the NMP corrective action program as CR-2014-001623. The NRC Resident Inspector has been notified.
The NMP Unit 2 Division 3 diesel testing was completed satisfactorily. The licensee declared the Division 3 diesel and HPCS systems operable at 1013 EST on 2/27/14 and exited TS 3.8.1 and 3.5.1. The licensee has notified the NRC Resident Inspector. Notified the R1DO (DeFrancisco). |
| ENS 49973 | 29 March 2014 21:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Division 3 Core Standby Cooling System Ventilation Failed | This report is being made pursuant to SAF 1.8, 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 steady state operations on Unit 1 at 1620 (CDT) hrs. on 3/29/14, the Division 3 Core Standby Cooling System (CSCS) Pump Room, SWGR Room, and Battery Room Ventilation failed in such a manner that heat could not be removed from the rooms. These Division 3 systems supply power and cooling water to the High Pressure Core Spray system (HPCS), which is a single-train system. The HPCS system and its associated power supplies were declared inoperable based on long-term temperature considerations. The system remains available due to manual damper adjustments that than can be made per an approved procedure. The licensee will notify the NRC Resident Inspector. |
| ENS 50296 | 22 July 2014 17:43:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Diesel Generator for High Pressure Core Spray Declared Inoperable During Surveillance Testing | This report is being made pursuant to 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. During the conduct of the Unit 2 Division 3 High Pressure Core Spray (HPCS) Diesel Generator (DG) surveillance test, one of 2 Cooling Water Outlet Valves failed to automatically open. The Division 3 Diesel is supplied by two redundant trains of cooling water one from each Service Water Divisional Header. Although the redundant cooling water supply was fully available and supplied adequate cooling to the diesel generator, the DG was at reduced margin to have adequate cooling water supply, if required during a loss of offsite power. Due to this loss of margin and inoperable condition, it has been determined that this failure could potentially affect the safety function of this system, and is being reported as an 8 hour ENS notification. The licensee has attributed the failure to high resistance in a relay which is currently being replaced. This places Unit 2 in the Technical Specification Action Statement 3.8.1, which requires restoration of Diesel Generator within 72 hours or commence a Reactor Shutdown. All other ECCS Systems have been verified operable. The licensee informed the NRC Resident Inspector and will inform the State of New York.
This update retracts Event Notification #50296, which reported an event or condition that could have potentially prevented fulfillment of a Safety Function needed to mitigate the consequences of an accident. Upon further review, it was determined that the ability of the HPCS system (single supported train) remained operable and capable of performing its safety function as evaluated by the NMP Unit 2 Safety Function Determination Process (TS 5.5.11). The NRC Resident Inspector has been notified. Notified the R1DO (Ferdas). |
| ENS 50337 | 2 August 2014 02:42:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable Due to Test Return Valve Leakage | River Bend Station personnel declared the High Pressure Core Spray system inoperable at 2142 (CDT) on 8/1/2014. The High Pressure Core Spray (HPCS) system at River Bend Station includes a test return line to the Condensate Storage Tank (CST). The test return line is isolated by two motor operated valves (MOVs) with both having a safety function to close on an ECCS (Emergency Core Cooling System) initiation signal to ensure that injection flow is directed to the reactor vessel. While the HPCS pump is normally aligned to the CST, the credited source of water for the pump is the suppression pool. Accordingly, the pump suction is realigned to the suppression pool on low level in the CST or when suppression pool level rises to a certain point. Station personnel identified leakage past the test return valves to the CST. In evaluating this condition, engineering personnel noted that the observed leakage past the two MOVs might be sufficient to deplete suppression pool inventory such that it would not be capable of performing its specified function for the duration of the 30 day mission time. The issue of concern is that once HPCS is aligned to the suppression pool post-LOCA, pool inventory would be lost to the CST through the leaking test return valves. Based on that concern, the HPCS pump suction valve from the suppression pool was disabled in the closed position to preserve pool inventory. This action caused the HPCS system to be declared inoperable at 2142 (CDT). This action results in a 14 day shutdown LCO and is reportable to the NRC in accordance with 10 CFR 50.72(b)(3)(v)(D). The HPCS pump remains available with its suction aligned to the CST. Assuming normal makeup water supplies are available, the HPCS system can be realigned to the suppression pool if necessary. This condition continues to be evaluated and rework options are being developed. The NRC Senior Resident Inspector has been notified. |
| ENS 50463 | 17 September 2014 00:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Loss of Division 3 of Shutdown Service Water Requires Hpcs to Be Declared Inoperable | At 1905 hours (CDT), during surveillance testing of the Division 3 Shutdown Service Water (SX) system, the Division 3 SX pump tripped for unknown reasons. The Division 3 SX system was declared inoperable and in accordance with Technical Specification 3.7.2, Action A, the High Pressure Core Spray (HPCS) system was declared inoperable. Since the HPCS system is a single train safety system, this event is reportable under 10CFR50.72(b)(3)(v)(D). An investigation is underway to determine the cause of SX pump to trip. The NRC Resident (Inspector) has been notified. |
| ENS 50551 | 20 October 2014 06:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Scram on Loss of Feedwater | The Perry Power Plant experienced a reactor scram during a shift of non-essential vital power supply to the alternate source. Feedwater was lost resulting in receiving a valid level 3 and level 2 signal. High Pressure Core Spray and Reactor Core Isolation Cooling started and injected. Reactor level and pressure have been stabilized to required bands. The motor feed pump has been started and is controlling level. High Pressure Core Spray and Reactor Core Isolation Cooling have been returned to standby. During the scram, all rods fully inserted into the core. Decay heat is being removed via the steam dumps to the condenser. The electrical grid is stable and supplying plant loads. An emergency diesel generator started, as designed, as a result of the level 2 signal but did not load. No safety valves lifted as a result of the transient. The cause of the loss of feedwater is under investigation. The licensee will be notifying the State of Ohio and Perry Township and has notified the NRC Resident Inspector.
The plant is currently in Mode 3, stable with cooldown and depressurization to Mode 4 in progress. Level control is being provided by the motor feedwater pump. Troubleshooting of the cause of the scram and loss of feed water is on-going. The initial notification identified 10CFR50.72(b)(3)(iv)(A), 'Specified System Actuation', as a reporting criteria. The specific system that actuated was not provided. As a result of receiving a reactor vessel water level 2 signal a containment/BOP isolation signal was received. All systems isolated as required and the plant is restoring isolated systems in accordance with procedure. The licensee will be notifying the State of Ohio and Perry Township and has notified the NRC Resident Inspector. Notified R3DO (Pelke). |
| ENS 50601 | 7 November 2014 13:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Scram Due to Loss of Feedwater | The Perry Nuclear Power Plant experienced an automatic reactor scram due to a loss of feedwater, which resulted in receiving valid reactor vessel water Level 3 and Level 2 initiation signals. The High Pressure Core Spray system and the Reactor Core Isolation Cooling system started and injected. Reactor water level and pressure have been stabilized in the required bands. The motor feed pump automatically started and is being used to control reactor vessel water level. The High Pressure Core Spray and Reactor Core Isolation Cooling systems have been returned to the standby mode. As a result of receiving a reactor vessel water Level 2 signal a Balance of Plant containment isolation signal was received. All systems isolated as required and the plant is restoring isolated systems in accordance with plant procedures. During the scram, all rods fully inserted into the core. Decay heat is being removed via turbine bypass valves to the main condenser. The electrical grid is stable and is supplying plant loads. An emergency diesel generator (Division 3 High Pressure Core Spray) started, as designed, as a result of the reactor vessel water Level 2 signal. No safety relief valves lifted as a result of the transient. The plant is stable with cooldown and depressurization to Mode 4 in progress. The cause of the loss of feedwater is under investigation. The NRC Resident Inspector has been notified. The State of Ohio and local officials will be notified. |
| ENS 50707 | 30 December 2014 05:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable Due to Leak on Division 3 Diesel Generator Cooling Water Pump | This report is being made pursuant to 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. During routine surveillance testing of the Unit 2 Division 3 Emergency Diesel Generator (LOS-DG-M3) a small pinhole leak was identified in the pump casing of the Division 3 Diesel Generator Cooling Water Pump. This condition has been evaluated and the Division 3 Diesel Generator Cooling Water Pump has been declared inoperable. The Division 3 Diesel Generator Cooling Water Pump 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 12/29/14 at 2330 (CST) when the HPCS system was determined to be inoperable. This condition could have prevented the High Pressure Core Spray System (HPCS), a single train safety system, from performing its design function. The licensee notified the NRC Resident Inspector. |
| ENS 51028 | 1 May 2015 04:44:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray System Inoperable | River Bend Station personnel declared the High Pressure Core Spray (HPCS) system inoperable at 2344 CDT on 4/30/2015. The HPCS system at River Bend Station includes a test return line to the Condensate Storage Tank (CST). The test return line is isolated by two motor operated valves (E22-MOVF010 and E22-MOVF011), with both having a safety function to close on an ECCS initiation signal to ensure that injection flow is directed to the reactor vessel. There is currently a blind flange installed downstream of these two valves. While the HPCS pump is normally aligned to the CST, the credited source of water for the pump is the suppression pool. Accordingly, the pump suction is realigned to the suppression pool on low level in the CST or when suppression pool level rises to a certain point. While performing maintenance on the downstream test return valve (E22-MOVF011), station personnel identified leakage past the upstream test return valve (E22-MOVF010) which was being used as an isolation boundary. In evaluating this condition, engineering personnel noted that the observed leakage past the upstream isolation MOV might be sufficient to deplete suppression pool inventory such that it would not be capable of performing its specified function for the duration of the 30-day mission time. The issue of concern is that once HPCS is aligned to the suppression pool post-LOCA, pool inventory would be lost due to the leaking upstream isolation valve (E22-MOVF010) and out the disassembled downstream isolation valve (E22-MOVF011). Based on that concern, the HPCS pump suction valve from the suppression pool was disabled in the closed position to preserve pool inventory. This action caused the HPCS system to be declared inoperable at 2344 CDT. This action results in a 14 day shutdown LCO and is reportable to the NRC in accordance with 10CFR50.72(b)(3)(v)D. The HPCS pump remained available with its suction aligned to the CST. Message has been left with NRC Senior Resident Inspector.
The licensee is retracting the report for Event No. 51028. On April 28, the High Pressure Core Spray System (HPCS) was inoperable to support planned maintenance. During repairs on the HPCS pump test return valves, leakage through the upstream isolation valve was observed when the downstream valve was disassembled. At 2315 (CDT) on April 30, it was conservatively determined that the leakage represented a potential challenge to the 30-day inventory of the suppression pool, and the pool was declared inoperable. At 2344 (CDT) on April 30, the HPCS pump suction valve to the suppression pool was closed to isolate that potential leakage path until the maintenance could be completed. This action returned the suppression pool to an operable status. On June 24, a quantitative leak rate test was performed on the upstream isolation valve (E22-MOVF010). That test determined that the leakage through the valve was not of such magnitude to have had the potential to deplete the 30-day inventory of the suppression pool during post-accident operation of the HPCS system. Additionally, when the HPCS pump suction valve on the suppression pool was closed on April 30, the system was already in a planned outage that commenced on April 28. As such, this condition need not have been reported. The licensee notified the NRC Resident Inspector. Notified the R4DO (Campbell). |
| ENS 51159 | 16 June 2015 08:52:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Emergency Diesel Generator Found Inoperable | At 0452 EDT hours on June 16, 2015, during performance of a surveillance test for the Division 3 4160 Volt Bus Undervoltage/Degraded Voltage Channel Calibration and Logic System Functional Test, the K36 degraded voltage time delay relay was found outside of the Technical Specification 3.3.8.1 allowable value, resulting in an inoperable condition of the Division 3 Emergency Diesel Generator (EDG). The Division 3 EDG had previously been declared inoperable for performance of the surveillance testing. The K36 degraded voltage time delay relay initiates load shedding, isolates the Division 3 bus, and starts the Division 3 EDG. The Technical Specification allowable value is 180 to 270 seconds. The as-found time was 272.66 seconds. The K36 relay was calibrated in accordance with plant procedures and returned to service at 0730 hours on June 16, 2015. The Division 3 EDG is the on-site power source for the High Pressure Core Spray system which is a single train system. Therefore, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been informed. |
| ENS 51201 | 6 July 2015 21:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Multiple Spurious Operations Scenario That Could Adversely Impact Post-Fire Safe Shutdown | A 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 51545 | 18 November 2015 05:55:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable | At 2355 (CST) on 11/17/2015, River Bend Station declared the High Pressure Core Spray (HPCS) system inoperable in accordance with Technical Specification 3.8.9, Condition E (Declare HPCS system and Standby Service Water System Pump 2C inoperable immediately) due to Division 1 Control Room Air Conditioning System HVK-CHL1C being inoperable because of a significant Freon leak on SWP-PVY32C. Actions taken to exit LCO: Alternated divisions of Control Room Air Conditioning System to Division 2 HVK-CHL1D in service and Division 1 HVK-CHL1A in standby. The basis for declaring High Pressure Core Spray inoperable was that the control room chiller also chills the switchgear room that supplies power to the HPCS. HPCS was out of service for less than one hour while the chillers were swapped from Division 1 to Division 2. The licensee has notified the NRC Resident Inspector. |
| ENS 51552 | 19 November 2015 13:24:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable | At 0724 (CST) on 11-19-2015, River Bend Station declared the High Pressure Core Spray System inoperable in accordance with Technical Specification 3.8.9, Condition E (Declare High Pressure Core Spray System and Standby Service Water System Pump 2C inoperable immediately) due to Division 2 Control Room Air Conditioning System Chiller HVK-CHL1D being inoperable due to a significant lube oil leak. HVK-CHL1D tripped on Low Lube Oil Differential Pressure. Division 1 Control Building Air Conditioning System Standby Chiller HVK-CHL1A automatically started as expected. Actions taken to exit LCO (Limiting Condition of Operation): Operators alternated to HVK-CHL1B in standby. The licensee notified the NRC Resident Inspector. |
| ENS 51600 | 11 December 2015 10:16:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable | At 0416 (CST) on 12-11-2015, River Bend Station declared the High Pressure Core Spray system INOPERABLE in accordance with Technical Specification 3.8.9, Condition E (Declare High Pressure Core Spray System and Standby Service Water System Pump 2C inoperable immediately) due to Division 2 Control Room Air Conditioning System HVK-CHL1D tripping off because of high inboard bearing temperature of 180 deg F. Actions taken to exit LCO: Alternated divisions of Control Room Air Conditioning System to Division 1 HVK-CHL1C in service and Division 2 HVK-CHL1B in standby and exited LCO at 0439. The licensee has notified the NRC Resident Inspector. |
| ENS 51800 | 17 March 2016 20:15:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Engineered Safety Feature System Actuation Due to Lightning Strike | At 1515 (CDT) on March 17th 2016, Grand Gulf Nuclear Station received a valid actuation signal of the Division 2 Engineered Safety Feature (ESF) Load Shedding and Sequencing system. The actuation signal was most likely caused by a lightning strike to the offsite power source supplying this ESF bus. This caused a loss of the in service shutdown cooling system and associated system actuations. Grand Gulf Nuclear Station (GGNS) was in Mode 5 at 85 (degrees) F coolant temperature. Reactor Cavity was flooded to High Water Level with a time to reach 200 (degrees) F of 7.5hrs. GGNS is conducting a planned refuel outage with core alterations in progress. Systems were aligned as follows: Division 2 Diesel Generator was OPERABLE and the associated ESF bus aligned to transformer ESF12 (115KV Port Gibson offsite feeder). Residual Heat Removal (RHR) system 'B' was in service in shutdown cooling being supplied from this ESF bus (16AB) with Alternate Decay Heat Removal available as a backup. Division 3 Diesel Generator was unavailable due to planned maintenance on support systems. The associated ESF bus was also aligned to transformer ESF12 (115KV Port Gibson offsite feeder). Division 1 Diesel Generator was available and the associated ESF bus aligned to the transformer ESF 11 (Switchyard offsite power feeders Baxter-Wilson and Franklin). Power was never lost to this bus. RHR 'A' and Low Pressure Core Spray (LPCS) were not available due to planned maintenance (tagged out of service). ESF 21 Transformer was out of service for planned maintenance. A suspected lightning strike caused a momentary perturbation in power in the 115KV Port Gibson line causing the Division 2 Load Shedding and Sequencing (LSS) system to actuate. This actuation caused a loss of Residual Heat Removal system 'B' due to being shed (expected). The Division 2 Diesel Generator started and tied onto the bus as expected, restoring power in 7 seconds. Shutdown cooling was restored at 1518 (CDT) and was out of service for 3 mins 13 sec. Reactor coolant and spent fuel pool temperatures remained at 85 (degrees) F throughout this scenario. Core Alterations were suspended and fuel placed in its designated location per the approved movement plan. Division 3 systems; High Pressure Core Spray, Standby Service Water System 'C', and Division 3 Diesel Generator were tagged out of service for planned maintenance. Division 3 Diesel Generator received a valid actuation signal but did not start due to being out of service. The Division 3 bus was restored manually to ESF 11. All safety systems operated as expected for the loss of power to ESF 12 and Division 2 LSS actuation. This is being reported under: 1. 10CFR50.72(b)(3)(iv)(A)-Specified system actuation; Division 2 LSS and Division 3 Diesel Generator start logic. 2. 10CFR50.72(b)(3)(v)(B)-RHR Capability; Loss of shutdown cooling. The NRC Resident Inspector has been notified. |
| ENS 51899 | 3 May 2016 03:29:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray System Inoperable Due to a Control Room Chiller Trip | At 2229 (CDT) on 05-02-2016, River Bend Station declared the High Pressure Core Spray system INOPERABLE in accordance with Technical Specification 3.8.9, Condition E (Declare High Pressure Core Spray System and Standby Service Water System Pump 2C inoperable immediately) due to Division 1 Control Room Air Conditioning System HVK-CHL1C being INOPERABLE due to a trip of the chiller on high inboard bearing temperature. Actions taken to exit the LCO: Alternated divisions of Control Room Air Conditioning System to Division 2 HVK-CHL1D in service and Division 1 HVK-CHL1A in standby. The licensee notified the NRC Resident Inspector.
Supplement: An operability evaluation has been performed based on system operating procedures in place at the time of this event, and on calculations regarding heat-up rates of the spaces served by the main control room air conditioning system. Operating procedures already in place on May 2 specified the operator actions required to restore the air conditioning system to service following the unanticipated trip of a chiller. The normal shift complement was on duty at the time of the event, and could have provided an adequate number of operators to accomplish this task. The operability evaluation made no new assumptions regarding availability of operators. The manual actions to be performed for the start of an alternate chiller following a trip of an in-service chiller system have been determined to require 2.15 hours, based on ANSI 58.8 guidance. (ANSI/ANS 58.8, Time Response Design Criteria for Nuclear Safety Related Operator Actions, provides the industry guidance In this regard.) Calculations of building heat-up rates have demonstrated that the loss of the air conditioning system can be sustained for 19 hours before temperatures in the rooms containing the Division 3 electrical equipment that support operability of the HPCS system exceed their maximum allowable ambient value. Based on the conclusions of the operability evaluation, the trip of the 'C' HVK chiller on May 2 had no actual adverse effect on the ability of the electrical distribution systems in the main control building to support the safety function of the HPCS system. Event Notification No. 51899 is hereby withdrawn. The licensee has notified the NRC Resident Inspector. Notified R4DO (Rollins). |
| ENS 52442 | 18 December 2016 19:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Scram Due to Load Reject from Substation | On 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. |
| ENS 52443 | 19 December 2016 07:20:00 | 10 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.
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). |
| ENS 52510 | 26 January 2017 02:36:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray System Diesel Room Fan Motor Failure | On 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 52516 | 28 January 2017 00:08:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable | This notification is to report a loss of safety function in accordance with 10 CFR 50.72(b)(3)(v)(D). At approximately 1808 CST hours on Friday, January 27, 2017, the Grand Gulf Nuclear Station Unit 1 High Pressure Core Spray (HPCS) system was declared inoperable due to the trip of the HPCS Jockey Pump. At the time of discovery, Unit 1 was in Mode 2 and raising power in the source range to return to power operations. No other safety systems were inoperable at the time of this event. Investigation into the cause of the event is ongoing and the system will be returned to operational status prior to proceeding to Mode 1. The licensee has notified the NRC Resident Inspector. |
| ENS 52519 | 31 January 2017 01:08:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | One Division Emergency Diesel Generator Cooling Water System Declared Inoperable | This 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 52601 | 9 March 2017 09:19:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Relay Failure | On March 7, 2017, Division 2 Residual Heat Removal (RHR) system was inoperable due to a scheduled maintenance system outage window. At 2258 (CST), Operations identified a Division 1 Unit Substation Switchgear relay was cycling, which is part of the Division 1 AC Power system. The specific relay could not be identified at the time. Division 1 AC Power systems were protected. On March 8, 2017 at 1830 hours, Division 2 RHR was restored to operable status. On March 9, 2017 at 0319 hours, Operations declared Division 1 Emergency Diesel Generator (EDG) inoperable due to the (identification of the) Division 1 relay as related to properly tripping non-essential loads on a bus under-voltage condition. The relay would not have actuated to trip non-essential loads. The proper tripping of non-essential loads is a requirement for Division 1 EDG. The Updated Safety Analysis Report (USAR) Emergency Core Cooling Systems (ECCS) analysis specifies with the Division 1 DG failure, the remaining systems available are: Automatic Depressurization System (ADS), High Pressure Core Spray (HPCS), and 2 Low Pressure Core Injection (LPCI) systems. As a result of Division 2 RHR (being) inoperable at the same time Division 1 EDG was inoperable, an unanalyzed condition existed. While Division 2 RHR was inoperable, Division 1 EDG was inoperable. Technical Specification (TS) Limiting Condition of Operation (LCO) 3.8.1, AC Sources - Operating, was not met. Condition B, One Required DG Inoperable, Required Action B.2 declares required features, (normally) supported by the inoperable DG, inoperable when the redundant required features are inoperable, with a completion time of 4 hours. The action would have required declaring Division 1 ECCS inoperable, which includes Division 1 RHR and Low Pressure Core Spray (LPCS). With Division 1 EDG, Division 1 RHR, and Division 2 RHR inoperable, the station did not satisfy the USAR ECCS analysis and was in an unanalyzed condition. This condition is reportable under 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition, since the condition occurred within three years of the date of discovery. The NRC Resident Inspector has been notified. |
| ENS 52631 | 23 March 2017 07:56:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable | River Bend Station personnel declared the High Pressure Core Spray (HPCS) system inoperable at 0256 on 3/23/2017. During performance of the HPCS Pump and Valve Operability Test, the operators observed an unusual system response after E22-MOVF023 (HPCS Test Return to the Suppression Pool) was stroked closed. A field check showed that the key that connects the E22-MOVF023 valve stem to the anti-rotation device had become dislodged. E22-MOVF023 is a Primary Containment Isolation Valve (PCIV) and is designed to close automatically on an ECCS (Emergency Core Cooling System) initiation signal to ensure that injection flow is directed to the reactor vessel. Technical Specification (TS) 3.6.1.3 requires that containment penetrations associated with an inoperable PCIV be isolated. E22-MOVF023 was declared inoperable at 0028. Operators were unable to close or demonstrate that E22-MOVF023 was fully closed as required by TS 3.6.1.3 and proceeded to isolate the associated containment penetration by closing other system valves. This action was completed at 0320. The net effect of the actions taken to isolate the containment penetration is that HPCS is inoperable as of 0256. This results in 14 day LCO. The licensee has notified the NRC Resident Inspector.
The Event Time was 0028 CDT rather than 0256 CDT. "The scheduled surveillance test of the high pressure core spray system was initiated at 2355 CDT on March 22, and the pump was secured at 0028 CDT on March 23. The inspection of the HPCS test return valve to the suppression pool occurred at 0050 CDT, and it was at that point that an apparent malfunction of the valve had occurred to the extent that it did not appear to be able to perform its safety function to close upon receipt of a design basis system initiation signal. Thus, the event time for this condition would be more accurately defined as 0028 CDT. Notified R4DO (James Drake) via e-mail. |
| ENS 52806 | 15 June 2017 14:58:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable | At 0958 hours (CDT), during planned surveillance testing of the Division 3 Shutdown Service Water (SX) subsystem, the Division 3 SX pump tripped for unknown reasons. The Division 3 SX subsystem was declared inoperable and in accordance with Technical Specification 3.7.2, Action A.1, the High Pressure Core Spray (HPCS) system was declared inoperable. Since the HPCS system is a single train safety system, this event is reportable under 10CFR50.72(b)(3)(v)(D). An investigation is underway to determine the cause of the SX pump trip. The NRC Resident has been notified. |
| ENS 52848 | 11 July 2017 14:50:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Flow 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.
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 52869 | 25 May 2017 23:47:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid High Pressure Core Spray (Hpcs) System Actuation in Mode 5 | At 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 52891 | 8 August 2017 19:54:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray System Suppression Pool Level Instrumentation Inoperable | On August 8, 2017, at 1554 hours (EDT), during restoration from testing of the High Pressure Core Spray (HPCS) Suppression Pool Level High Instrumentation, unexpected as-left indications were found that impacted both of the required channels of instrumentation. Subsequent venting of the instrumentation lines was completed and both channels of instrumentation are reading consistent with previously taken as-found data. The instrumentation was declared OPERABLE at 1635. The initial cause of the unexpected as-left indications appears to be the introduction of air into the instrumentation lines during the calibration activities. This is considered a loss of safety function based on both of the HPCS Suppression Pool Level High Instrumentation channels being declared INOPERABLE and the loss of the automatic HPCS suction swap to the Suppression Pool on a high level. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The (NRC Resident Inspector) has been notified. |
| ENS 53000 | 4 October 2017 06:50:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Perry Commenced Technical Specification Required Shutdown | On October 4, 2017, at 0250 hours (EDT), the Perry Nuclear Power Plant commenced a Technical Specification (TS) shutdown by lowering reactor power from 100 percent rated thermal power to 98 percent to comply with TS LCO 3.0.3. Reactor power was further reduced to 82 percent rated thermal power at 0430 hours (EDT). The plant had entered TS 3.0.3 at 0155 hours (EDT) upon loss of MCC (Motor Control Center), Switchgear, and Miscellaneous Electrical Equipment Areas HVAC System train A while train B was removed from service for maintenance. MCC switchgear ventilation train A was declared inoperable based on excessive belt noise and a dropped belt on MCC switchgear supply fan A. This also constitutes a loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(2)(i) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector was notified.
Due to the loss of both trains of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC, actions were taken in LCO 3.8.7 for AC and DC Distribution Systems, LCO 3.8.4 for DC Sources, LCO 3.8.1 for AC Sources, and the associated support systems, the High Pressure Core Spray system was also declared inoperable, which is a single train safety system and therefore, an additional loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B), 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D). At 0620 hours (EDT) the A train of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC and High Pressure Core Spray was declared operable and LCO 3.0.3 was exited. The plant was restored to 100% (percent) power at 0804 (EDT). The NRC Resident Inspector was notified. Notified R3DO(Hills). |
| ENS 53110 | 9 December 2017 19:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Manual Reactor Scram Due to Loss of Division 1 Ac Power to Numerous Components | At approximately 1347 (CST) on 12/09/17, the Main Control Room received annunciators that indicated a trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1 breaker. Numerous Division 1 components lost power (powered from unit subs 1A and A1). The Division 1 containment Instrument Air isolation valves had failed closed by design due to the loss of power. Due to the loss of containment instrument air, several control rods began to drift into the core as expected and, by procedure, the reactor mode switch was placed in the shutdown position at 1353 (CST). All control rods fully inserted. Also due to the loss of power, the Fuel Building ventilation dampers failed closed by design. With the normal ventilation system secured, secondary containment differential pressure rose to slightly greater than 0 inches water gauge which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge at 1348 (CST). The Control Room entered EOP-8, Secondary Containment Control. Secondary Containment differential pressure was restored within Technical Specification requirements at 1351 (CST) by starting the Division 2 Standby Gas Treatment system. This event is being reported as a manual actuation of the Reactor Protection System (RPS) and as a Condition that Could Have Prevented Fulfillment of a Safety Function. The cause is currently under investigation. The NRC Resident has been notified. The licensee informed the NRC Resident Inspector.
During a review of plant logs it was identified that the primary to secondary containment differential pressure was identified to be outside of Technical Specification 3.6.1.4 limits of 0 plus or minus 0.25 psid at 2009 on 12/9/17 due to the primary containment ventilation system dampers closing as a result of the loss of power. This parameter is an initial safety analysis assumption to ensure that primary containment pressures remain within the design values during a Loss of Coolant Accident (LOCA). As a result, this condition is reportable as an unanalyzed condition that significantly degrades plant safety. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).
During the post transient review of the trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1, it was identified that the unplanned INOPERABILITY of the Low Pressure Core Spray (LPCS) system due to the loss of power to the injection valve constitutes an event or condition that could have prevented fulfillment of a safety function and is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. The High Pressure Core Spray (HPCS) remained available to perform the core spray function, if necessary, during a design basis Loss of Coolant Accident (LOCA), however HPCS and LPCS are each considered single train safety systems. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone). |
| ENS 53117 | 12 December 2017 19:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Declared Inoperable | At approximately 1330 CST on Tuesday, December 12, 2017, Grand Gulf Nuclear Station declared Division 3 'C' Battery inoperable due to questions concerning battery terminal connection continuity. Technical Specification 3.8.4, DC Sources - Operating, Condition E, Required Action E.1, requires the station to declare the High Pressure Core Spray System inoperable immediately. The Division 3 'C' Battery and High Pressure Core Spray System was declared operable and the LCOs (Limiting condition of operation) were declared met at 1731CST on Tuesday, December 12, 2017. Based on field measurements of terminal torque and resistance, the as-found and as-left terminal resistance micro-ohm readings indicated satisfactorily all times. Formal evaluation of the as-found condition of the battery is in progress. This report is to notify the NRC of a loss of safety function on the High Pressure Core Spray System. The NRC Resident Inspector was notified. |
| ENS 53137 | 23 December 2017 04:48:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray System Declared Inoperable | High Pressure Core Spray System was declared inoperable due to the discovery of a through-wall leak on the Minimum Flow line. Leak rate is 60 drops per minute from ASME Class 2 Piping. The leak has been isolated and the High Pressure (Core Spray) System has been placed in Secured Status. High Pressure Core Spray is considered a single train safety system. Inoperability of (the) High Pressure Core Spray System is considered an 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. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Senior Resident Inspector was notified. Technical Specifications Limiting Condition for Operation 3.5.1 Condition B was entered, requiring restoration of the High Pressure Core Spray System in 14 days. The licensee plans to notify State and Local Governments (Lake, Geauga, and Ashtabula Counties). |
| ENS 53219 | 17 February 2018 08:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Damaged Bus Bar Identified Potentially Affecting High Pressure Core Spray | This 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 53365 | 26 April 2018 20:31:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Inadvertent Injection of High Pressure Core Spray | River Bend Station experienced an inadvertent initiation and injection of High Pressure Core Spray (HPCS) at 1531 (CDT) on 4/26/2018 while operating at 100 percent power. During replacement of Level Transmitter B21-LTN081C 'Reactor Vessel Low Water Level 1', Main Control Room received an inadvertent initiation and injection of High Pressure Core Spray. The HPCS injection valve was open for approximately 40 seconds before the operators manually closed the valve. Feedwater Level Control responded per design and maintained Reactor Water Level nominal values. The Division 3 Diesel Generator (DG) also automatically started in response to the actuation signal. The DG did not automatically connect to the Division 3 switchgear since there was not a low voltage condition on the bus. The manual closure of the injection isolation valve caused the system to be incapable of responding to an automatic actuation signal. The manual override of the injection isolation valve was reset approximately 16 minutes after the event, restoring the system to its standby condition. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a condition that caused ECCS (Emergency Core Cooling System) discharge to RCS (Reactor Coolant System) and 10 CFR 50.72(b)(3)(v)(D) as a condition that caused the loss of function of the HPCS System. The Senior NRC Resident inspector has been notified. |
| ENS 53765 | 2 December 2018 05:00:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Alert Declared Due to Emergency Diesel Generator Damage | During the post-maintenance testing run of the Division III Emergency Diesel Generator (EDG), (a field operator) reported smoke coming from the diesel and an emergency shutdown was required. After the EDG was shutdown, significant damage (thrown rod) to the EDG was observed. Emergency Action Level HA 2.1 (an Alert) was declared at 0530 (EST). Currently, the plant is stable and operating at 100 percent power. All safety systems are available. The damage occurred approximately 20 minutes into the required 1 hour run. The licensee's emergency response organization has been activated. No offsite assistance was required or requested. There is a 14-day shutdown limiting condition for operation (LCO) in effect under technical specification 3.5.1 for the high pressure core spray system. Notified DHS Senior Watch Officer, FEMA Operations Center, DHS NICC Watch Officer, HHS Operations Center, DOE Operations Center, EPA Emergency Operations Center, FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email). The licensee has notified state and local authorities and the NRC Resident Inspector.
The licensee terminated the Alert at 0731 EST on 12/2/18. The basis for termination was that the licensee has met all procedural requirements to terminate the emergency and on-shift personnel can operate the unit without further assistance. Notified R1DO (Burritt), NRR EO (Miller), IRD MOC (Gott), HQPAO (Couret), ERDS Activation Group, DHS Senior Watch Officer, FEMA Operations Center, DHS NICC Watch Officer, HHS Operations Center, DOE Operations Center, EPA Emergency Operations Center, FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email). |
| ENS 53788 | 12 December 2018 06:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | En Revision Imported Date 12/17/2018 | EN Revision Text: MANUAL REACTOR SCRAM DUE TO FAILED OPEN TURBINE BYPASS VALVE At 1351 CST, the reactor was manually shutdown due to 'A' Turbine Bypass Valve opening. The Main Steam Line Isolation Valves were manually closed to facilitate reactor pressure control. Reactor level is being maintained through the use of Reactor Core Isolation Cooling System, Control Rod Drive System, and High Pressure Core Spray System. High Pressure Core Spray System was manually started to initially support reactor water level control. Reactor Pressure is being controlled through the use of the Safety Relief Valves and the Reactor Core Isolation Cooling System. The plant is stable in MODE 3. The cause of the 'A' Turbine Bypass Valve opening is under investigation at this time. The NRC Resident Inspector has been notified.
This is an update to EN # 53788 to correct an error on the event classification block of the form. The original notification did not have the block for 8 hour notification for Specified System Actuation checked. The actuation of Reactor Core Isolation Cooling System was discussed in original notification. The licensee notified the NRC Resident Inspector. Notified R4DO (Taylor). |
| ENS 53824 | 13 January 2019 14:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | En Revision Imported Date 3/8/2019 | EN Revision Text: HIGH PRESSURE CORE SPRAY SELF TEST FAILURE On January 13, 2019, the Self Test System reported a fault associated with the logic system for the High Pressure Core Spray (HPCS) high reactor water level closure function that could prevent the system from performing its safety function. The HPCS system was subsequently declared inoperable with actions taken per LCO (Limiting Condition for Operation) 3.6.1.3 to close and deactivate the 1E12-F004 valve, a primary containment isolation valve. Since HPCS is an emergency core cooling system and is a single train safety system, this condition is reportable under 10 CFR 50.72(b)(3)(v)(D) 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.' The NRC Resident Inspector has been notified. HPCS is in a 14-day technical specification LCO action statement.
Testing of the logic system load driver card for the High Pressure Core Spray (HPCS) high reactor water level closure function was completed both on site and at General Electric Hitachi (GEH). This testing determined the cause of the self-test system fault report was limited to the self-test portion of the load driver card and did not impact the ability of HPCS system to perform its specified safety function. Based on the testing results, this event is not reportable under 10 CFR 50.72(b)(3)(v)(D), 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.' Therefore, EN 53824 is being retracted. The NRC Resident Inspector has been notified. Notified the R3DO (Hills). |
| ENS 54197 | 3 August 2019 07:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | En Revision Imported Date 8/7/2019 | EN Revision Text: AUTOMATIC REACTOR SCRAM ON LOW REACTOR WATER LEVEL At 0226 (CDT), an automatic scram on low reactor water level occurred due to a trip of the 'B' Reactor Feed pump. All control rods fully inserted. Reactor water level 2 was reached and the High Pressure Core Spray system, Reactor Core Isolation Cooling system, Division 3 diesel generator, Standby Gas Treatment Systems 'A' and 'B' and all shutdown safety related service water pumps started as expected. Reactor Core Isolation Cooling and High Pressure Core Spray injected as expected. All level 2 containment isolation signals occurred as expected and all level 2 containment valves closed as expected. Reactor water level is currently being controlled in band by condensate. Reactor pressure is being maintained by main turbine Bypass Valves. This event is being reported under 10 CFR 50.72(b)(2)(iv)(A), for ECCS discharge to RCS; 10 CFR 50.72(b)(2)(iv)(B), for RPS actuation, and 10 CFR 50.72(b)(3)(iv)(A), for specified system actuation. The NRC Senior Resident Inspector has been notified. No safety relief valves lifted during the transient. The plant is in a normal shutdown electrical lineup with all safety equipment available. The licensee notified the Illinois Emergency Management Agency per their communications protocol.
Following automatic initiation of the High Pressure Core Spray (HPCS) System as described above, the HPCS System was manually secured following station procedures after verification that additional RPV (reactor pressure vessel) injection was no longer required. Securing HPCS injection in this manner prevents automatic restart of the system in the event of a subsequent low RPV level condition, rendering it inoperable. As the HPCS system is considered a single train safety system, this meets the reportability requirements of 10 CFR 50.72(b)(3)(v)(D). This reportable condition was identified following review of post-scram actions. The HPCS system has been restored to a Standby lineup. The licensee will be notifying the NRC Resident Inspector. Notified R3DO (Pelke).
Following the scram, the Primary Containment to Secondary Containment and the Drywell to Primary Containment differential pressure limits were exceeded. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.4, Primary Containment Pressure, and 3.6.5.4, Drywell Pressure, Actions A.1, B.1, and B.2 were entered. Primary Containment to Secondary Containment differential pressure and Drywell to Primary Containment differential pressure were restored to within the LCO limits at 1505 on 8/3/19 and the associated TS Actions were exited. This event is reportable under 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that could have prevented the fulfillment of the primary containment function due to being outside the initial conditions to ensure that drywell and containment pressures remain within design values during a loss of coolant accident. This event is also reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of the drywell and primary containment functions to control the release of radioactive material for the same reason. The licensee notified the NRC Resident Inspector. Notified R3DO (Pelke). |
| ENS 54244 | 28 August 2019 18:16:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Loss of High Presssure Core Spray | On Wednesday, August 28, 2019, at 1316 CDT, Grand Gulf Nuclear Station experienced a power loss to the Control Room High Pressure Core Spray (HPCS) Instrumentation Panel due to an internal inverter failure. The power loss caused the loss of the HPCS System (a single train system). The minimum flow valve (a Primary Containment Isolation Valve) for HPCS opened due to this power loss as well. This valve was manually closed in response to this, and the outboard isolation requirement for the associated penetration (which) is closed (for the) system remained intact throughout this event. No other accident mitigation systems were affected by this event. The cause of this event is under investigation at this time. The NRC Resident Inspectors were notified. This Condition is an 8-hour reportable condition as an event or condition that could have prevented the fulfillment of a safety function, in accordance with 10 CFR 50.72(b)(3)(v)(D). |
| ENS 54292 | 25 September 2019 06:38:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Loss of High Pressure Core Spray System | At 2338 PDT on September 24, 2019, the High Pressure Core Spray (HPCS) system was declared inoperable due to a leak on DSA-PCV-2C (2 inch Diesel Starting Air Pressure Control Valve). With one of two air headers isolated and being drained for maintenance, this leak caused the remaining starting air header for HPCS-GEN-DG3 (HPCS Diesel Generator) to lower to less than the operability limit. Upon declaring the HPCS system 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. 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 per 10 CFR 50.72(b)(3)(v)(D). The HPCS system is a single train system at Columbia. The leak was isolated and starting air header pressure restored to the HPCS diesel generator at 0104 PDT on September 25, 2019, and all associated Technical Specifications were exited. The NRC Resident Inspector was notified. |
| ENS 54364 | 1 November 2019 07:16:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray (Hpcs) System Inoperable | On November 1, 2019 at 0316 EDT, Nine Mile Point Unit 2 (NMP2) received Control Room annunciation for HPCS SYSTEM INOPERABLE and inoperable status light indication for TRIP UNITS OUT OF FILE/POWER FAIL. Initial investigation has identified a potential failed 24 vdc power supply which supplies power to the HPCS trip units for system initiation and control. The HPCS system has been declared inoperable per TS 3.5.1 resulting in an unplanned 14 day LCO. All other plant systems functioned as required. NMP2 is currently at 100 percent power in Mode 1. This condition is reportable under 10 CFR 50.72(b)(3)(v)(D) as, '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: (D) Mitigate the consequences of an accident.' The licensee notified the NRC Resident Inspector. |
| ENS 54371 | 6 November 2019 00:11:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Two Diesel Generators Concurrently Inoperable | On November 5, 2019 at 1811 CST, station service water A and the Division 1 diesel generator (DG) were declared inoperable based on the results of an engineering evaluation of a Class 3 piping leak. This was determined to be a potential inability to fulfill a safety function due to concurrent inoperability of two emergency diesel generators. Division 3 DG was inoperable due to planned maintenance on November 4, 2019 at 0000 CST. This event is being reported an 8-hour non-emergency notification per 10 CFR 50.72 (b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of a safety function (Accident Mitigation). Division 3 DG and high pressure core spray have been restored, and the fulfillment of the accident mitigation safety function has been restored. The NRC Senior Resident Inspector has been notified.
This was initially reported under 10 CFR 50.72(b)(3)(v)(D). However, subsequent engineering evaluation determined that the condition did not affect safety system operability. The evaluation determined that the leakage was within allowable limits and piping structural integrity was not challenged at this time nor in the past three years. The Division 1 DG and SSW A were at the time of discovery OPERABLE and EN54371 is being retracted. The licensee will notify the NRC Resident Inspector. Notified R4DO (Drake). |
| ENS 54403 | 21 November 2019 18:25:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | EN Revision Imported Date : 1/13/2020 UNIT 1 HIGH PRESSURE CORE SPRAY INOPERABLE On 11/21/2019, at 1225 CST, as a result of Division 4 DC bus voltage oscillations, bus voltage lowered to less than the required improved technical specification (ITS) voltage of 127.6 VDC. This resulted in declaring High Pressure Core Spray (HPCS) system inoperable per technical specification LCO 3.8.4 and 3.8.9 actions. Division 4 DC bus voltage was restored to greater than 127.6 VDC at 1227 CST. The HPCS system remains inoperable due to Division 4 DC battery charger inoperability. Since HPCS is an emergency core cooling system and is a single train safety system, this condition is reportable under 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified. Clinton Power Station has implemented required compensatory actions due to the Division 4 DC battery charger and HPCS remaining inoperable.
The purpose of this notification is to retract a previous report made on 11/21/2019 (EN 54403) under 10 CFR 50.72(b)(3)(v)(D). Subsequent to the initial notification, the event and the NRC guidance in NUREG-1022 pertaining to 10 CFR 50.72(b)(3)(v)(D) were reviewed further. The evaluation determined that the Division 4 DC bus voltage oscillations were caused by a degraded but operable charger. The Division 4 battery remained fully charged during the event and its operability was not impacted. Therefore, the HPCS system remained Operable. Under these circumstances, this event does not represent an inoperability of an accident mitigation system under 10 CFR 50.72(b)(3)(v)(D). Therefore, EN 54403 is retracted. The NRC Resident Inspector has been notified. Notified R3DO (Hanna). | |
| ENS 54693 | 4 May 2020 21:40:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Diesel Generator Cooling Water System Declared Inoperable | This 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. A through wall leak was found on piping connected to the Division 3 Diesel Generator (DG) Cooling Water Strainer. This condition has been evaluated and the Division 3 DG Cooling Water System has been declared inoperable. The Division 3 DG Cooling Water System is a support system for the Division 3 Emergency DG and the High Pressure Core Spray System (HPCS). The NRC Resident Inspector has been notified.
This update retracts Event Notification #54693, which reported a condition that could have potentially prevented fulfillment of a safety function needed to mitigate the consequences of an accident. An evaluation of the flaw on the piping connected to the Unit 2 Division 3 Diesel Generator (DG) Cooling Water strainer concluded that the system would have remained operable. The High Pressure Core Spray system, supported by the operable DG Cooling Water system, remained operable and capable of performing its safety function. The NRC Resident Inspector has been notified. Notified R3DO (Stone). |
| ENS 54880 | 4 September 2020 01:48:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Inoperable | River Bend Station experienced an inadvertent initiation and injection of High Pressure Core Spray (HPCS) at 2048 (CDT) on 9/3/2020 while operating at 92% power. Initial investigation indicates a power supply failure in the Division III trip units which feeds HPCS and Division III Diesel Initiation signals. The Control Room Operator responded to the event by taking manual control of Feedwater Level Control to maintain Reactor Water Level nominal values. The HPCS injection valve was open for approximately 25 seconds before operators manually closed the valve. The manual closure of the injection isolation valve caused the system to be incapable of responding to an automatic actuation signal. The manual override of the injection isolation valve was reset approximately 52 minutes after the event. The HPCS system has remained inoperable. The event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that caused loss of function of the HPCS System. No radiological releases have occurred due to this event. The Senior NRC Resident Inspector has been notified. These conditions put the unit in a 14-day LCO (3.5.1) for HPCS Inoperability and a 30-day LCO (3.7.1) for one Standby Service Water Pump Inoperable (2C). |
| ENS 55046 | 23 December 2020 12:53:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Inoperable | At 0653 CST on 12/23/20, it was discovered the single train of high pressure core spray was inoperable. Due to this inoperability, the system was in a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). All other emergency core cooling systems were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The high pressure core spray is inoperable because the water lake pump tripped. This inoperability puts the licensee in a 14-day limiting condition for operability. |
| ENS 55451 | 9 September 2021 05:33:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Inoperable | At 0033 CDT on September 9, 2021, Grand Gulf Nuclear Station (GGNS) was operating at 70 percent power when the High Pressure Core Spray (HPCS) was declared inoperable. The inoperability determination was made due to control room annunciations. In accordance with GGNS Technical Specification 3.5.1.B.1, the Reactor Core Isolation Cooling system was verified to be operable. Troubleshooting is in progress. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as an event or condition which could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified. |
| ENS 55682 | 1 January 2022 17:10:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray System Declared Inoperable | The Licensee provided the following information via fax: During performance of a surveillance of the High Pressure Core Spray (HPCS) service water system on January 1, 2022, the HPCS system was declared inoperable for performance of the surveillance. During the surveillance, pump discharge pressure and flow were above the action range curve specified in the surveillance. For the given flow rate, pump discharge pressure was too high. This condition prevents declaring the HPCS service water system and HPCS system operable. The HPCS service water and HPCS systems remain inoperable. The station entered Technical Specification (TS) 3.7.2.A and TS 3.5.1.B at 0910 (PST) on January 1, 2022. In accordance with TS 3.5.1.B, the Reactor Core Isolation Cooling (RCIC) system was verified to be operable. TS 3.5.1 Action B provides a 14-day completion time to restore HPCS to an operable status. All other Emergency Core Cooling systems (ECCS) are operable. 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 per 10 CFR 50.72(b)(3)(v)(D). The HPCS system is a single train system at Columbia. The NRC resident has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee is investigating the cause of the high pump discharge pressure and verifying instrumentation accuracy.
This Notification is to retract EN 55682, Unplanned High Pressure Core Spray (HPCS) Inoperability. On 1/1/2022 at (1735 EST), Columbia Generating Station notified the NRC under 10 CPR 50.72(b)(3)(v)(D) of the inoperability of a single train of safety system (HPCS) for performance of the surveillance. During the surveillance pump discharge pressure and flow were above the action range curve specified in the surveillance. Engineering performed an analysis of this event and concluded the HPCS was operable during the event and would have performed its required safety function. The results of initial IST testing of HPCS-P-2 via OSP-SW/IST-Q703 on 01/01/22 resulted in measured parameters falling outside of the acceptable range specified for this pump. Systematic error was suspected as the cause of the failure and the test was reperformed following taking actions to eliminate the suspected systematic errors. The second performance of the test on 01/01/22 resulted in acceptable pump performance. Evidence exists that the initial performance of the test failed due to imprecise averaging techniques due to difficulties in averaging continuously changing values on the test instrument. The second performance of OSP-SW/IST-Q703 should be considered a successful test and the test of record as the systematic error was eliminated and measured parameters are considered valid. The NRC Resident Inspector has been notified. The HOO notified R4DO (Rolando-Otero). |