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| Start date | Report date | Site | Reporting criterion | System | Event description | |
|---|---|---|---|---|---|---|
| ENS 56928 | 18 January 2024 20:04:00 | Columbia | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Service water Residual Heat Removal Spray Pond | The following information was provided by the licensee via email: On January 18, 2024, at 0030 PST, diesel generator 2 (DG2) was shut down following a monthly surveillance run. Subsequently, a leak was discovered in the DG2 building. Service water pump '1B' was secured at 0117, effectively stopping the leak. The leak was determined to be service water coming from a diesel generator mixed air cooling coil. Service water system 'B' and DG2 were subsequently declared inoperable at 0135. After discussion with engineering, it was identified that the amount of service water leakage from the cooling coil was assumed to be greater than the leakage allowed by the calculation to assure adequate water in the ultimate heat sink to meet the required mission time of 30 days. At 1204, it was determined that entry into Technical Specification 3.7.1 condition D was warranted since the assumed leakage from the cooling coil could exceed the calculated allowed value. At 1238, the control power fuses for service water pump '1B' were removed. DG2 and service water system 'B' were declared unavailable, and the technical specification condition for the inoperable ultimate heat sink was exited. With the control power fuses removed, the pump is kept from auto starting, effectively preventing the leak and ensuring the safety function of the ultimate heat sink is maintained while the cooling coil is repaired or replaced. Due to the leakage assumed greater than the calculated allowable value this condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition and per 10 CFR 50.72(b)(3)(v)(B) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to remove residual heat. There was no impact to the health and safety of the public. The NRC Resident has been notified.
The following information was provided by the licensee via email: On January 18, 2024 at 2138 EST, Columbia Generating Station notified the NRC under 10 CFR 50.72(b)(3)(ii)(B) of an unanalyzed condition on the available capacity of the ultimate heat sink (UHS) and under 10 CFR 50.72(b)(3)(v)(B) of an event or condition that could have prevented fulfillment of the safety function of structures or systems needed to remove residual heat. On January 18, 2024, following monthly surveillance of the diesel generator DG2, a DG2 room cooler flow alarm was received at 0115. A leak was discovered in the diesel mixed air (DMA) air handler unit. Service Water Pump '1B' was secured and the leakage was stopped at 0117. The service water system 'B' and diesel generator system 'B' were declared inoperable at 0135. The leak was assumed to be greater than that allowed to ensure adequate water in the UHS required to meet the 30-day mission time, and the UHS was declared inoperable at 1204. Control power fuses for the service water pump '1B' were removed to fully eliminate the leakage path from the cooler, and the UHS was declared operable at 1238. Following the event, engineering performed an analysis based on the size and location of the leak, and concluded it would have taken 1.4 days to deplete the available excess water in the UHS to below the minimum technical specification required water level of the spray pond. Operations were able to secure the service water subsystem of the UHS prior to exceeding the volumetric margins in the spray ponds to ensure the 30-day mission time was met. The condition did not represent a safety significant unanalyzed condition nor a loss of safety function. The NRC Resident Inspector has been notified. Notified R4DO (Gepford). | |
| ENS 56838 | 7 November 2023 17:00:00 | Seabrook | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Steam Generator Feedwater Auxiliary Feedwater | The following information was provided by the licensee via email: On November 07, 2023 at 1200 EST, it was discovered that all pumps in the Auxiliary Feedwater system were inoperable due to the loss of control power to the 'B' train Emergency Feedwater (EFW) flow control valve which supplies the 'D' steam generator. The redundant 'A' train EFW control valve for the 'D' steam generator remains functional, as well as the capability of the Auxiliary Feedwater system to supply all steam generators. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(B). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The 'A' and 'B' EFW Flow Control Valves are arranged in a series configuration for each Steam Generator. Failure of any of the 8 EFW Flow Control Valves to meet its Surveillance Requirements will render all EFW Pumps inoperable per tech specs. | |
| ENS 56620 | 12 July 2023 08:49:00 | Millstone | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Service water Emergency Core Cooling System | The following information was provided by the licensee via email: At 0449 (EDT) on 7/12/2023, Millstone Unit 3 declared the 'B' train of the emergency core cooling system (ECCS) inoperable due to a degraded damper associated with the ventilation support system for the 'B' charging pump. At the time of this event, the 'A' train of service water was already inoperable due to planned maintenance on a breaker that would have prevented an 'A' service water valve powered from this breaker from closing on a safety signal. This configuration resulted in the possibility that the 'A' train of ECCS would not have been available to fulfill its design function under all postulated accident conditions. This event is being reported under 10 CFR 50.72(b)(3)(v)(B), '(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: (B) remove residual heat).' Subsequently, the 'A' train of service water was restored to operable at 0548 on 7/12/2023. Repairs and investigation continue on the 'B' train ECCS damper. The NRC resident has been notified. This event did not impact Millstone Unit 2. There was no impact to the public.
The following information was provided by the licensee via email: The condition was reported to the NRC pursuant to 10 CFR 50.72(b)(3)(v)(B), via an 8-hour report as 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 remove residual heat. A subsequent engineering review of the conditions that existed at the time determined that, based on area temperature response, any impact on ventilation flows into and out of the `B' charging pump cubicle did not generate an observable change in the temperature trend. Based on this, it is concluded with reasonable assurance that the functional requirement of the support system was maintained and the `B' charging pump would have continued to perform its safety function until the `A' train of service water was restored to operable and as a result safety function was not lost. Therefore, this condition is not reportable and NRC Event Number 56620 is being retracted. The basis for this conclusion has been provided to the NRC Resident Inspector." Notified R1DO (Bicket). | |
| ENS 56569 | 11 June 2023 05:30:00 | Beaver Valley | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | The following information was provided by the licensee via email: At 0130 EDT on June 11, 2023, it was discovered that the Beaver Valley Power Station, Unit No. 2 auxiliary building door A-35-5A, credited for tornado missile protection of the primary component cooling water system, was open and unlatched. Upon discovery, the door was shut and latched. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | ||
| ENS 56429 | 23 March 2023 15:45:00 | Fermi | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | The following information was provided by the licensee via fax: While in Mode 1 at 100 percent power at 1145 EDT on March 23, 2023, it was determined that all mechanical draft cooling Tower (MDCT) fan brakes would not perform their design function during a tornado due to a design flaw with the control system. The MDCT fan brakes are required to prevent fan overspeed from a design basis tornado. The MDCT fans are required to support the operability of the ultimate heat sink (UHS). At the time of discovery, the provisions of LCO 3.0.9 were being utilized for loss of the 'D' MDCT fan brake (barrier loss). When it was identified the condition was a design flaw common to all MDCT fan brakes, the 24-hour allowance for restoration was entered. A design change is currently being implemented to restore MDCT fan brake operability. This condition is reportable per 10 CFR 50.72(b)(3)(v)(A), (B), & (D). The NRC Resident Inspector has been notified. | ||
| ENS 56023 | 30 July 2022 03:17:00 | Dresden | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | The following information was provided by the licensee via email: At 2217 CDT on 7/29/22, cribhouse suction bay levels were reported less than 501.5 feet due to buildup of grass on bar racks. Ultimate Heat Sink (UHS) is INOPERABLE due to Surveillance Requirement 3.7.3.1 not met. ENTER Technical Specification (TS) 3.7.3 condition A (Required Action (RA) A.1 mode 3 in 12 hours, RA A.2 mode 4 in 36 hours). Dresden Lockmaster reports river level normal at 504.89 feet. Commenced trash rake operations to clear grass debris off of intake bar racks. At 0135 CDT on 7/30/22, cribhouse suction bay levels were reported at greater than 501.5 feet. Exit TS 3.7.3 condition A. 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)(B). There was no impact on the health and safety of the public or plant personnel. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee notified the NRC Resident Inspector.
At 1116 CDT on 7/30/22, a repeat condition occurred where cribhouse suction bay levels were reported < 501.5 feet due to buildup of grass on bar racks. Entered TS 3.7.3 condition A (RA A.1 mode 3 in 12 hours, RA A.2 mode 4 in 36 hours). Actions are in-progress to clear grass debris off the intake bar racks. 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)(B). At 1745 CDT on 7/30/22, cribhouse suction bay levels were reported at >501.5 feet. Exit TS 3.7.3 condition A. The station continues to monitor for intake grass buildup and taking appropriate actions to maintain UHS operability. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee notified the NRC Resident Inspector. Notified R3DO (Peterson). | ||
| ENS 55891 | 10 May 2022 18:59:00 | Quad Cities | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Core Spray Low Pressure Coolant Injection | The following information was provided by the licensee via fax: At 1359 CDT on May 10, 2022, the 1B LPCI Loop Upstream Injection valve (1-1001-28B) was found to have a motor operated torque switch issue and inadequate lubrication. This issue called into question the ability of the valve to close when required. At 1746 CDT on May 10, 2022, both trains of Unit 1 LPCI were made simultaneously inoperable. TS 3.6.1.3 Condition A required de-activation of 1B LPCI Loop Downstream Injection valve (1-1001-29B) which was completed at 1746 CDT. Because of the de-activation of the 1B LPCI Loop downstream injection valve and LPCI Loop select logic, both trains of LPCI were made 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). Unit 1 HPCI and both loops of Core Spray are operable. After further engineering review, it was determined that 1B LPCI Loop Upstream injection valve condition was minor in nature and would not have affected the ability of the valve to close when required. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via phone call and email: The last sentence in the second paragraph, "After further engineering review, it was determined that 1B LPCI Loop Upstream injection valve condition was minor in nature and would not have affected the ability of the valve to close when required," has been deleted. The licensee is continuing to follow up on the issue and believes that sentence to be unclear and premature. Notified R3DO (Skokowski). | |
| ENS 55831 | 7 April 2022 23:00:00 | Turkey Point | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Residual Heat Removal | The following information was provided by the licensee via fax or email: At 1900 EDT on 04/07/22, while Unit 4 was in Mode 4 following a refueling outage, it was discovered that both trains of residual heat removal (RHR) were simultaneously inoperable due to gas voiding. At 2032 EDT corrective actions were completed and both trains of RHR were declared operable. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(B). The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: On 4/8/2022 at 0110 EDT Turkey Point Unit 4 notified the (NRC Operations Center (NRCOC)) pursuant to 10 CFR 50.72(b)(3)(v)(B) that both trains of Residual Heat Removal (RHR) were simultaneously inoperable due to the presence of gas voids that were identified during scheduled system gas accumulation testing. Subsequent evaluation by (Florida Power & Light (FPL)) Engineering has concluded that both trains of RHR remained operable and capable of performing their specified safety function. This NRCOC notification is a retraction of EN# 55831. The licensee notified the NRC Resident Inspector. Notified R2DO (Miller) | |
| ENS 55612 | 27 November 2021 10:19:00 | Oconee | 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 | Decay Heat Removal | At 0519 EST on November 27, 2021, with Unit 2 in Mode 5 at zero percent power, an actuation of the Emergency AC Electrical Power System occurred. The reason for the Emergency AC Electrical Power System auto-start was a lockout of the CT-2 transformer; causing a temporary loss of AC power to the main feeder bus. The Keowee Hydroelectric Units 1 and 2 automatically started as designed when a main feeder bus undervoltage signal was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Emergency AC Electrical Power System. Additionally, the temporary loss of AC power resulted in a loss of Decay Heat Removal (DHR) that was restored upon power restoration to the main feeder bus. Therefore, this condition is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v) for an event or condition that could have prevented fulfillment of a safety function. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The loss of the CT-2 transformer is under investigation. Main feeder bus power was restored within a minute so no plant heat up occurred as a result of the loss of the decay heat removal system. | |
| ENS 55492 | 25 September 2021 07:42:00 | Nine Mile Point | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Reactor Protection System | On 9/25/2021 at 0342 EDT, Nine Mile Point Nuclear Station Unit 1 experienced a loss of the Uninterruptible Power Supply (UPS), UPS 162A, which resulted in a loss of Reactor Protection System (RPS) Bus 11. This resulted in an isolation of both No. 11 and No. 12 Emergency Condensers. Emergency Condenser No. 11 was returned to standby on 9/25/2021 at 0420 EDT and Emergency Condenser No. 12 was returned to standby on 9/25/2021 at 0429 EDT. This 8-hour non-emergency report is being made based upon requirements of 10 CFR 50.72(b)(3)(v)(B) which states: 'Licensee shall notify the NRC of 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: (B) remove residual heat.' The NRC Senior Resident was informed. | |
| ENS 54425 | 5 December 2019 14:10:00 | Cooper | 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)(3)(ii)(B), Unanalyzed Condition | The following was received via email from Cooper Nuclear Station: At 0810 (CST), on 12/5/19, Operations personnel discovered BLDG-DOOR-R209, FIRE DOOR BETWEEN CRITICAL SWITCHGEAR ROOMS F & G, was unlatched. The door was immediately latched upon discovery. Based on door logs, the door separating the two critical switchgear rooms was inadvertently left unlatched for approximately 5 minutes. This door is a Steam Exclusion Boundary (SEB) door. It is required to be closed and latched when the Auxiliary Steam Boiler is in service due to Auxiliary Steam piping passing through Critical Switchgear Room 'G'. If a steam line break was to occur with the door unlatched, steam could render both Critical Switchgear busses inoperable. This is being reported under 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition, and 10 CFR 50.72(b)(3)(v), 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 (B) remove residual heat and to (D) mitigate consequences of an accident. There was no impact on the health and safety of the public or plant personnel. The door closes automatically and appeared to have been left unlatched by the last person passing through. The door was tested and latches as required. The licensee notified the NRC Resident Inspector. | ||
| ENS 54229 | 19 August 2019 13:24:00 | Davis Besse | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Feedwater Auxiliary Feedwater | At 0924 EDT, on August 19, 2019, it was discovered that both trains of the Auxiliary Feedwater System were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(B). The door separating the two steam-driven Auxiliary Feedwater Pump Rooms was inadvertently left open during maintenance activities for more than an hour. The door was immediately closed upon discovery, restoring operability to the Auxiliary Feedwater System. The non-safety grade Motor-Driven Feedwater Pump remained operable during this time; additionally, the beyond-design basis diesel-driven Emergency Feedwater Pump also remained available. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 54286 | 6 July 2019 05:56:00 | Catawba | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(v)(A, B, and D) - Event or Condition that Could Have Prevented the Fulfillment of a Safety Function Unit 2 Component Cooling water system inoperable. On July 6, 2019, from 0156 to 1545 (EDT), it was determined that both trains of the unit 2 component cooling water system were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The affected safety function was restored on July 6, 2019, at 1545 (EDT) when the 2B component cooling train was restored to operable. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. At the time of the event no other safety related systems were inoperable. The event had no impact on Unit 1. | ||
| ENS 54078 | 22 May 2019 06:56:00 | Susquehanna | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 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 | HVAC | On 5/22/2019, the 'A' Control Structure Chiller (Div I) tripped due to a loss of (motor control center) MCC 0B136. The 'B' Control Structure Chiller was already inoperable due to Div II (Emergency Service Water) ESW being out of service for planned maintenance. With the loss of Control Structure HVAC System the ability to maintain temperatures in various spaces including relay rooms, Control Room Floor Cooling and Emergency Switchgear rooms was lost. The 'B' Control Structure Chiller was restarted at 0251 EDT and cooling was reestablished to the required areas, however the 'B' chiller is not considered operable at this time. Units 1 and 2 entered (Technical Specification) TS 3.0.3 at 0256 EDT and a controlled shutdown of both units commenced, Unit 2 at 0340 EDT and Unit 1 0350 EDT. This constitutes a TS required shutdown and requires a 4 hour (Emergency Notification System) ENS notification in accordance with 10 CFR 50.72(b)(2)(i). The failure also requires an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(v) due to the loss of a safety function. The licensee needs to restore the 'B' loop of ESW to exit the Limiting Condition of Operation (LCO). The licensee is currently performing a flow surveillance, once complete and assuming the data is acceptable, the licensee will be able to exit the LCO. The units are in a normal electrical lineup. The licensee will be notifying the state of Pennsylvania FEMA Operations Center. The licensee has notified the NRC Resident Inspector.
On 5/22/2019 at 0601 EDT Susquehanna Steam Electric Station reported a shutdown had been commenced at 0340 EDT for Unit 2 and 0350 EDT for Unit 1 due to inoperability of both control structure chillers. Power has been restored to MCC 0B136, and at 0901 EDT the 'A' control structure chiller was declared operable and LCO 3.0.3 was exited. Power reduction for both units was halted at 0901 EDT and preparations for power restoration initiated. As of 1255 EDT on 5/22/2019, Unit 1 power is 94% and Unit 2 power is 92%. Notified the R1DO (Arner). | |
| ENS 53759 | 28 November 2018 06:00:00 | Callaway | 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 | Main Steam Line | On November 28, 2018, while performing an engineering review of the bases for environmental qualification (EQ) requirements for the Atmospheric Steam Dumps (ASDs), it was determined that applicable EQ requirements had not been applied to a key component of each of the ASDs. The result of this issue is that it the availability of the ASDs for a controlled plant cooldown following a postulated steam line break outside containment cannot be assured. Callaway is developing a compensatory action temporary plant modification to install insulation that will protect the affected ASD components from the post Main Steam Line Break temperature. This condition is reportable 10 CFR 50.72(b)(3)(v) for 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 (B) remove residual heat, or (D) mitigate the consequences of an accident. The issue places the plant in a 24-hour Technical Specification (TS) Limiting Condition for Operations (LCO), 3.7.4. The licensee has notified the NRC Resident Inspector. | |
| ENS 53699 | 28 October 2018 04:00:00 | Indian Point | 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 | Service water | During the performance of Service Water Essential header swap, SWN-6 (Supply to Turbine Building Oil Coolers) valve stem became disconnected from its gear box at 85% open and could not be operated. Therefore, the non-essential service water system was inoperable. LCO 3.0.3 was entered at 0930 (EDT) with required actions to be in Mode 3 in 7 hours, Mode 4 in 13 hours and Mode 5 in 37 hours. Repair efforts were successful at shutting SWN-6, and LCO 3.0.3 was exited at 1305 (EDT) before adding any negative reactivity in support of shutdown. ('TS Required S/D' box not checked.) This condition constituted a loss of safety function which requires an 8 hour report (in accordance with) IAW 50.72(b)(3)(v)(B): Without the ability to close SWN-6, the non-seismic portion of the conventional Service Water System could not be isolated as required in the event of either a seismic event or as required in the EOPs. The nonessential service water system is required to support the recirculation phase post (Design Basis Accident) DBA for accident mitigation. The licensee notified the NRC Resident Inspector and the State of New York. | |
| ENS 53693 | 24 October 2018 05:00:00 | Quad Cities | 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 | Emergency Diesel Generator Core Spray Low Pressure Coolant Injection | On October 24, 2018 at 0901 CDT, during performance of the 'Functional Test of Unit 1 Second Level Undervoltage,' a loss of Bus 13-1 and Bus 18 occurred. The 1/2 Emergency Diesel Generator (EDG) automatically started due to a valid actuation on loss of power to Bus 13-1, but did not load due to required testing alignment. The loss of Bus 13-1 caused the loss of the 1A loop of Core Spray, both loops of Low Pressure Coolant Injection (LPCI), and Bus 18. All equipment responded as expected. Bus 13-1 and Bus 18 were restored at 0911(CDT) on 10/24/18. Other affected systems are in the process of being restored. An investigation as to the cause of the event has been initiated. This notification is being made in accordance with 10 CFR 50.72(b)(3)(iv), 'Event or Condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B),' because the 1/2 EDG auto started due to the loss of power condition. This notification is also being made in accordance with 10 CFR 50.72(b)(3)(v)(B), 'Event or Condition that Could Have Prevented Fulfillment of a Safety Function,' because both loops of LPCI were inoperable for a short time period. During the ten minutes where LPCI was unavailable, Unit 1 was in Technical Specification LCO 3.0.3. Unit 1 is currently in LCO 3.8.1(b) until the EDG is restored. Unit 2 was not affected by this event. The licensee will notify the NRC Resident Inspector. | |
| ENS 53546 | 10 August 2018 05:00:00 | Prairie Island | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | Emergency Diesel Generator | On 8/10/2018 at 1445 (CDT) both trains of Cooling Water (Cooling Water Pumps for Emergency Diesel Generators) were declared INOPERABLE and both units entered (Technical Specification) (TS) 3.0.3 due to corroded jacket cooling water plugs for (the) 12 and 22 cooling water pump motors; therefore this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). At 1543 (CDT), 08/10/2018 the 121 Cooling Water pump was aligned to the "A" Cooling Water train and the TS 3.0.3 condition was exited for both units. (After restoring train A cooling water the site entered a seven day limiting condition for operations, TS 3.7.8 for one inoperable cooling water pump.) There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
Testing and forensic analysis performed subsequent to the notification has determined the as-found condition would not have impacted either diesel-driven pumps' ability to start, run, and meet flow/pressure requirements to perform their required safety function. Therefore, EN# 53546 is being retracted. The NRC Resident Inspector has been notified of the Event Notification retraction. Notified R3DO (Kozak). | |
| ENS 53419 | 22 May 2018 04:00:00 | Beaver Valley | 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)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | EN Revision Text: GAS VOIDS DISCOVERED IN BOTH TRAINS OF LOW HEAD SAFETY INJECTION On 5/22/2018, while operating at approximately 100 percent power, Ultrasonic Testing of the Beaver Valley Power Station (BVPS) Unit 1 Low Head Safety Injection (LHSI) pump suction piping identified gas voids in excess of the acceptable limit for void volume. Both trains of LHSI were declared inoperable. Technical Specification (TS) 3.5.2 for both trains of the LHSI system was entered along with TS 3.0.3 which requires the initiation of a plant shutdown. Time of TS entry was 12:56 (EDT). Plant shutdown was commenced at 15:56 (EDT) in accordance with plant procedures. At 15:59 (EDT) Train 'A' LHSI was restored to operable status, TS 3.0.3 Action was exited and the power reduction was stopped at approximately 99 percent. At 17:43 (EDT) Train 'B' LHSI was restored to operable status, TS 3.5.2 Actions were exited. This is reportable per 10 CFR 50.72(b)(3)(ii) Unanalyzed Condition, 10 CFR 50.72(b)(3)(v) Event or Condition that Could Have Prevented the Fulfillment of a Safety Function and 10 CFR 50. 72(b )(2)(i) TS Required Shutdown. The NRC Resident Inspector has been notified.
Further engineering evaluation has determined that the gas voids that existed at the time of discovery would not have rendered the LHSI (Low Head Safety Injection) system inoperable if it were required to actuate. The engineering evaluation concluded that filling of the containment sump during a Design Basis Accident would result in a void volume reduction such that the void in the LHSI suction piping would not be large enough to significantly impact the operability of the system. Therefore, the system remained operable but degraded. No TSs (Technical Specifications) were required to be entered and no shutdown was required. As such, all three reporting criteria do not apply and are being retracted. The NRC Resident Inspector has been notified. Notified R1DO (Burritt). | ||
| ENS 53388 | 7 May 2018 18:35:00 | Callaway | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | Steam Generator Reactor Coolant System Auxiliary Feedwater Residual Heat Removal Main Steam Line | On May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power. During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops. The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop. Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs. This condition is reportable per 10 CFR 50.72(b)(3)(v) for 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 (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident." The licensee notified the NRC Resident Inspector of this condition. | |
| ENS 53382 | 4 May 2018 16:29:00 | River Bend | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(ii)(B), Unanalyzed Condition | During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, River Bend Station identified non-conforming conditions in the plant design such that specific TS equipment is considered to not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Diesel Generator Building through conduit and pipe penetrations. A tornado could generate multiple missiles capable of striking Division 1, Division 2, and Division 3 Diesel Generator support equipment rendering all Safety Related Diesel Generators inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for 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 (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector. | ||
| ENS 53374 | 1 May 2018 20:51:00 | Grand Gulf | 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 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Shutdown Cooling Reactor Pressure Vessel Residual Heat Removal Emergency Core Cooling System | At 1551 hrs (CDT) on 5/1/2018, with the plant in Mode 5, a division one Reactor Pressure Vessel (RPV) Level 1 signal was received; however there was no actual change in RPV level. RPV Level remained at High Water Level supporting refuel operations. This caused an actuation of division one Load Shed and Sequencing system that shed and then re-energized the 15 bus. Division one diesel generator started from standby. Residual Heat Removal pump 'A', which was in shutdown cooling mode, was lost during the bus shed, and was re-sequenced upon re-energization of the 15 bus. Upon restoration of shutdown cooling, the RHR pump discharged into the RPV. RCS temperature increased approximately 5 degrees Fahrenheit as a result of the loss of shutdown cooling. The cause of the actuation signal is under investigation. In accordance with NUREG 1022, Event Reporting Guidelines, this event is conservatively reported under 10 CFR 50.72(b)(2)(iv)(A) as an event that results in emergency core cooling system discharge into the RCS as a result of a valid signal, under 10 CFR 50.72(b)(3)(iv)(B)(8) as an event that results in the actuation of emergency ac electrical power systems, and under 10 CFR 50.72(b)(3)(v)(B) as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function (remove residual heat). The licensee notified the NRC Resident Inspector. | |
| ENS 53356 | 22 April 2018 06:22:00 | Watts Bar | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Residual Heat Removal Emergency Core Cooling System | On April 22, 2018 at 0222 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 entered TS (Technical Specifications) LCO (Limiting Condition for Operation) 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 0227 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions. The NRC Resident Inspector has been notified.
This event is being retracted. The initial report was based on a conservative acceptance criteria for gas accumulation adopted on April 19, 2018 when it was determined that the previously used acceptance criteria for gas accumulation in the ECCS was non-conservative. Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. With a void of less than the acceptance criteria, in the event of ECCS actuation, the system piping support loads will remain within structural limits and the piping system will remain operable. Therefore, both trains of Unit 2 RHRS were operable and the previously reported 10 CFR 50.72(b)(3)(v)(B) event is being retracted. The NRC Resident Inspector staff has been informed of this event retraction. Notified R2DO (Desai) of this retraction. | |
| ENS 53355 | 22 April 2018 01:52:00 | Watts Bar | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Residual Heat Removal Emergency Core Cooling System | On April 21, 2018 at 2152 EDT, Watts Bar Nuclear Plant (WBN) Unit 1 entered TS (Technical Specifications) LCO (Limiting Condition for Operation) 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 2222 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions. The NRC Resident Inspector has been notified.
This event is being retracted. The initial report was based on a conservative acceptance criteria for gas accumulation adopted on April 19, 2018 when it was determined that the previously used acceptance criteria for gas accumulation in the ECCS was non-conservative. Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. With a void of less than the acceptance criteria, in the event of ECCS actuation, the system piping support loads will remain within structural limits and the piping system will remain operable. Therefore, both trains of Unit 1 RHRS were operable and the previously reported 10 CFR 50.72(b)(3)(v)(B) event is being retracted. The NRC Resident Inspector has been informed of this event retraction. Notified R2DO (Ehrhardt). | |
| ENS 53353 | 20 April 2018 15:42:00 | Braidwood | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Residual Heat Removal | On Friday, April 20, 2018 at 1042 CDT, Braidwood Station Unit 1 was at 0 percent power in Mode 6. The 1A Diesel Generator (DG) was inoperable with troubleshooting in progress. The 1B DG was being run for a normal monthly run in accordance with 1 BwOSR 3.8.1.2-2, 'Unit One 1B Diesel Generator Operability Surveillance,' and subsequently tripped. The trip was due to a failure of the overspeed butterfly valve actuator and springs, and not an actual overspeed condition. The unit entered Technical Specification (TS) 3.8.2, 'AC Sources - Shutdown,' Condition B for required DG inoperable. All required TS actions were met at the time of the 1B DG inoperability. The offsite power source remains available. At no time was residual heat removal lost. This event is reportable under 10 CFR 50.72(b)(3)(v)(B) for 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 remove residual heat. The licensee has notified the NRC Resident Inspector. | |
| ENS 53259 | 13 March 2018 14:00:00 | Pilgrim | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Reactor Protection System Emergency Diesel Generator | On March 13, 2018 at 1000 hours (EDT), with the reactor in Cold Shutdown condition, both 345kV incoming power lines and 23 kV Shutdown Transformer became unavailable during the Northeast winter storm. Per procedures, the emergency on-site emergency power supplies (Emergency Diesel Generators) were running and providing power to essential systems. In addition, the back-up Diesel Air Compressor was in service and one Reactor Protection System bus was on the back-up power supply prior to the loss. With both 345kV incoming power lines and 23 kV Shutdown Transformer unavailable, Pilgrim Nuclear Power Station procedures direct a report be made to the NRC per the requirements of Title 10 Code of Federal Regulations 50.72(b)(3)(v), any event that could have prevented the fulfillment of the safety function. No actual loss of safety function has occurred since the on-site emergency power supplies are maintaining the reactor in a safe shutdown condition and removing residual heat. The loss of incoming power is under investigation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. | |
| ENS 53223 | 20 February 2018 18:25:00 | Callaway | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(ii)(B), Unanalyzed Condition | Service water Auxiliary Feedwater | At 1225 CST, all three Auxiliary Feedwater (AFW) pumps were declared inoperable at the Callaway Plant upon discovery that a door (DSK13311) credited for protection of equipment from the effects of a high-energy line break (HELB) hazard had come partially open due to vibration harmonics from the running turbine-driven Auxiliary Feedwater pump (TDAFP). Immediate investigation identified that play in the mechanism that holds the door closed had rendered it susceptible to movement from the vibration harmonics. The affected HELB door specifically protects safety-related instruments that provide a swap-over signal upon detection of a low suction pressure condition for the AFW pumps and thereby automatically effect a suction transfer for the AFW pumps from the condensate storage tank (normal/standby source) to the Essential Service Water (ESW) system (credited safety-related source). All of the AFW pump suction transfer instrument channels were declared inoperable. Per Technical Specifications (TS) 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation,' the applicable Condition(s) and Required Action(s) for inoperable AFW pump suction transfer instrumentation only addresses a single channel being inoperable. Thus, the condition of having all three instrument channels inoperable required entry into TS Limiting Condition for Operation (LCO) 3.0.3. At the same time, however, with the automatic suction transfer capability rendered inoperable, all three AFW pumps, i.e., the TDAFP and the 'A' and 'B' motor-driven AFW pumps, were declared inoperable. Although LCO 3.0.3 was applicable, entry into the Required Actions of LCO 3.0.3 was suspended per the Note attached to Required Action E.1 of TS 3.7.5, 'Auxiliary Feedwater (AFW) System,' which states, 'LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.' At 1336 CST, Operations took actions to prevent the TDAFP from running, so the remaining (AFW Pumps) could be returned to Operable status. Operations then declared the affected instrumentation and the 'A' and 'B' motor-driven AFW pumps Operable. This allowed LCO 3.0.3 and Conditions A, B, D, and E under TS 3.7.5 to be exited. With only the TDAFP inoperable, TS 3.7.5 Condition C and its Required Actions remain in effect. Due to the degraded HELB door rendering all three AFW pumps inoperable, the unidentified condition is being reported as an unanalyzed condition that significantly degraded plant safety (per 10 CFR 50.72(b)(3)(ii)(B)) as well as a condition that could have prevented the fulfillment of the safety functions of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and mitigate the consequences of an accident (per 10 CFR 50.72(b)(3)(v)(A), (B), and (D), respectively). The NRC Senior Resident Inspector has been notified.
Event Notification (EN) # 53223, made on 2/20/2018, is being retracted because new information has been obtained that negates the original basis for reporting the unanalyzed condition. Specifically, an evaluation of the HELB that is postulated to occur in the TDAFP room has determined that without crediting door DSK13311 for protection, the affected safety-related instruments would not be exposed to environmental conditions beyond their analyzed capability. This resulted in a conclusion that the unanalyzed condition of door DSK13311 being open did not prevent the affected safety-related instruments or their supported AFW pumps from performing their required safety functions to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and/or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(A), (B), or (D). The licensee notified the NRC Resident Inspector. Notified R4DO (Drake). | |
| ENS 53000 | 4 October 2017 06:50:00 | Perry | 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 | HVAC High Pressure Core Spray | 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 52959 | 10 September 2017 16:51:00 | Browns Ferry | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | At 1151 (CDT) on September 10, 2017 Browns Ferry Units 1 and 2 declared 'B' Control Bay chiller inoperable. 'A' Control Bay chiller was previously declared inoperable. This resulted in inoperability of the equipment in the U1 and U2 4kV Shutdown Board Rooms. The declarations of the equipment in the Shutdown Board Rooms is a loss of safely function for electrical components (4kv Shutdown Boards and 480V Shutdown Boards) required for shut down of the U1 and U2 reactors and maintaining them in a safe shutdown condition, as well as RHR capability and Accident Mitigation. lnoperability of these boards also requires declaring two trains of Standby Gas Treatment inoperable resulting in a loss of safety function for Units 1, 2 and 3 for systems needed to control the release of radioactive material. This event requires an 8 hour report IAW 50.72(b)(3)(v), 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The NRC Resident Inspector has been notified. CR 1336821 was initiated in the Browns Ferry Corrective Action Program. | ||
| ENS 52836 | 30 June 2017 05:30:00 | Indian Point | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Auxiliary Feedwater | On June 30, 2017 at time 0130 (EDT), the Condensate Storage Tank (CST) was declared inoperable per Technical Specification 3.7.6. A pin hole sized through wall leak was discovered on the downstream side of CD-123, 32 Auxiliary Boiler Feed Pump Bearing Cooling Relief Valve, which is unisolable to the CST. No type of Non-Destructive Examination (NDE) can be used to effectively characterize the defect. Based on the inability to characterize the defect and the fact that the degradation mechanism is not readily apparent, the valve is considered inoperable because the valve has lost its ability to maintain a pressure boundary. As a result, the CST is inoperable. Currently, the CST has 580,000 gallons of water contained in the tank, which is well above the minimum required amount of 360,000 gallons. City water is the backup means to supply water to the Auxiliary Feedwater System, and this has been verified to be operable in accordance with the Actions of the Technical Specification. The CST must be restored to operable status within 7 days. The CST provides cooling water to remove decay heat and the minimum amount of water in the Condensate Storage Tank is the amount needed to maintain the plant for 24 hours at hot shutdown following a trip from full power. The CST satisfies Criteria 2 and 3 of 10 CFR 50.36. This event was determined to be reportable as a Loss of Safety Function pursuant to 10 CFR 50.72(b)(3)(v)(B). The licensee will notify the NRC Resident Inspector and the State of New York. | |
| ENS 52812 | 16 June 2017 20:35:00 | Oconee | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Service water | Keowee Hydro Units (KHU) 1 and 2 were both declared inoperable at 1635 (EDT) on 6-16-17 due to discovery of breaker 1GSC-1 (KHU-1) in the intermediate position, and breaker 2GSC-1 (KHU-2) in the open position. Keowee Hydro Units are required to be operable per TS (Technical Specification) 3.8.1 (AC Sources - Operating), TS 3.8.2 (AC Sources - Shutdown), and TS 3.7.10 (Protected Service Water, applies only to KHU aligned to the Overhead Power Path). All Tech Spec required conditions were entered, and all required actions completed. Both Standby Buses were energized from a Lee Combustion Turbine via an isolated power path at 1715 (EDT) on 6-16-17 in accordance with TS 3.8.1 Condition (I), Required Action (I.1). It has been determined by station personnel that a loss of safety function did occur between 1635 (EDT) (when the Keowee Hydro Units were declared inoperable) and 1715 (EDT) (when the Standby Buses were energized from a Lee Combustion Turbine via an isolated power path). Investigation has determined the cause of breakers 1GSC-1 and 2GSC-1 being out of their required closed position to be inadvertent bumping while performing station work activities. Breakers 1GSC-1 and 2GSC-1 have been reclosed, and both Keowee Hydro Units have been declared operable as of 2351 (EDT) on 6-16-17. The licensee notified the NRC Resident Inspector. | |
| ENS 52701 | 21 April 2017 06:02:00 | Davis Besse | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Emergency Diesel Generator | On 4/21/17, high grid voltage conditions were experienced, resulting in voltages higher than those established for operability of the offsite circuits. Grid voltages have been observed at approximately 355.8 kV on the nominal 345 kV system. As a result, both qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System were declared inoperable at 0202 hours (EDT). Main generator voltage control has been lowered to the minimum possible excitation with the unit operating at 100 percent power. The inoperability of both offsite circuits results in a loss of safety function in accordance with NRC reporting guidance. The voltage of the onsite Essential busses remains within acceptable values, and both Emergency Diesel Generators are operable. At 0715 (EDT) on 4/21/17, grid voltage has returned to an acceptable value and the equipment was declared operable. The NRC Resident Inspector has been notified.
Following the reporting of high switchyard voltage on 4/21/17, the licensee-established voltage limits were re-evaluated. The new high voltage limit has been established at 362.94 kV on the nominal 345 kV system, or 105.2 percent of nominal voltage, as compared to the previous maximum grid voltage of 103.3 percent. This new limit is above the 355.8 kV experienced on 4/21/17. Therefore, the equipment remained operable and no loss of safety function existed for the qualified circuits between the offsite transmission network and the onsite Class 1E AC electrical power distribution system, and the notification made per 10 CFR 50.72(b)(3)(v)(A-D) is being retracted. An evaluation of the past three years of switchyard voltage data was also performed, and it was concluded the AC power system and its connected safety-related equipment remained capable of performing its required safety functions during the three-year evaluation period. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. Notified the R3DO (Stone). | |
| ENS 52647 | 29 March 2017 19:08:00 | Palisades | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(ii)(B), Unanalyzed Condition | Feedwater Service water Emergency Diesel Generator Auxiliary Feedwater | During an evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Palisades Nuclear Plant personnel identified conditions in the plant design such that specific TS equipment is considered not adequately protected from tornado missiles. Specifically, vulnerabilities were identified in the following systems and components: Service Water System - Service water pump discharge header and service water pump cable trays. Fuel Oil Transfer System - Fuel oil transfer piping and transfer pump cable trays. Emergency Diesel Generators - Vent lines on the fuel oil day tanks. Control Room Heating, Ventilation, and Cooling System - Both the normal and emergency intake ducts. Steam Driven Auxiliary Feedwater Pump - Feedwater pump relief valves. Component Cooling Water System - Component cooling water surge tank. The identified vulnerabilities are being addressed in accordance with Enforcement Guidance Memorandum (EGM), 15-002, and Interim Staff Guidance, DSS-ISG-2016-01. Initial compensatory measures are in place. The licensee notified the NRC Resident Inspector. | |
| ENS 52607 | 13 March 2017 18:58:00 | Callaway | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(ii)(B), Unanalyzed Condition | Emergency Diesel Generator | On March 13, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Callaway Plant identified non-conforming conditions in the plant design such that specific Technical Specification equipment is considered to not be adequately protected from tornado missiles. The Emergency Fuel Oil Truck Connection Lines for both redundant Emergency Fuel Oil trains extend through the Plant South wall of the Diesel Generator Building structure where they may be exposed to design bases tornado missile impact. The direct impact by the design basis missile could result in damage to the fuel oil transfer lines, thereby preventing delivery of the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. This condition affects the fuel supply to both supported Emergency Diesel Generator trains. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for 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 (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. These conditions are being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The NRC Resident Inspector has been notified.
Event Notification EN # 52607, made on 03/13/2017, is being retracted because new information has been obtained that negates the originally reported condition. Specifically, subsequent to the Event Notification, an engineering analysis was performed which confirmed that a design basis missile strike on either of the unprotected truck connection lines would not result in damage to the extent that the affected fuel oil transfer lines would be prevented from delivering the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. (The analysis showed that although bending / deformation of the lines would occur in response to the postulated missile strike, integrity of the lines would remain.) Based on the above, the unanalyzed condition did not prevent the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition does not meet the criteria for 8-hour notification that are provided in 10 CFR 50.72(b)(3)(ii)(B) or 10 CFR 50.72(b)(3)(v)(A), (B), or (D). As the condition does not require enforcement discretion, the provisions of EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents) need not be invoked. However, the immediate and long-term compensatory actions that have been taken following discovery of the condition will remain in place until the condition is fully resolved. The NRC Resident Inspector has been informed of this Event Notification retraction. Notified R4DO (Azua). | |
| ENS 52566 | 18 February 2017 21:37:00 | River Bend | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Shutdown Cooling Reactor Recirculation Pump | At 1537 CST on February 18th, 2017, while the plant was in MODE 5 for a scheduled refueling outage, the main control room experienced a loss of Control Building chilled water and the associated ventilation systems while attempting to alternate divisions for testing. An equipment malfunction in a breaker supplying a Main Control Room air handling unit caused a loss of both divisions of Control Room and Control Building chilled water systems and associated ventilation systems until 1737 CST. During the period between 1537 and 1737, neither division of Control Building chilled water was able to perform the support function for cooling Division 1 and 2 AC and DC power distribution systems or the support function for the Division 1 and 2 Control Room Fresh Air systems. Shutdown Cooling remained in service throughout this event. There were no apparent effects on any plant equipment from the loss of chill water and ventilation. The Division 1 Control Building chill water and ventilation system was returned to service at 1737 on February 18, 2017. Actions were initiated to terminate the OPDRV (operations with potential to drain the reactor vessel) that was in progress at the time of the event by installing the reactor recirculation pump seal. As a conservative measure, actions were initiated to set containment and containment was set at 2145. Troubleshooting and analysis is ongoing to confirm and correct the problem which caused the loss of the Control Building chill water and ventilation system. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B). The NRC Senior Resident Inspector has been notified. | |
| ENS 52517 | 29 January 2017 08:09:00 | River Bend | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Shutdown Cooling | At 0209 CST, on January 29, 2017, while the plant was in MODE 4 for a refueling outage, the main control room crew removed the AC/DC inverter in the Division 1, 120 VAC electrical distribution system from service due to an equipment malfunction. Removing the inverter from service caused a loss of the associated 120 VAC instrument buss. This instrument buss loss caused a trip of the Division 1 Control Building Chill Water and Ventilation system. The Division 2 Control Building Chill Water and Ventilation System was locked out for surveillance testing at the time of the equipment failure. This condition rendered both divisions of Control Building Chill Water and Ventilation Systems unable to perform the support function for cooling Division 1 and 2 AC and DC power distribution systems. These systems are required to support the operability of two required divisions of shutdown cooling. Division 2 Shutdown Cooling System was in service and remained in service through out the event. The Division 2 Control Building Chill Water and Ventilation System was returned to service at 0220 CST on January 29, 2017. Division 1 Control Building Chill Water remains inoperable pending restoration with the installed backup Division 1 DC/AC inverter. Actions are ongoing to place this component in service and restore the associated 120 VAC instrument buss. The equipment malfunction was limited to the Division 1 inverter. The investigation of the inverter failure is ongoing. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B). The NRC Senior Resident Inspector has been notified. | |
| ENS 52369 | 16 November 2016 15:45:00 | Susquehanna | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Emergency Diesel Generator | On November 16, 2016 at 1045 (EST), Unit 2 B ESS (Engineered Safeguards System) Bus Sync Selector switch failed and was unable to be switched out of the ON position. The failure rendered all other sync selector switches associated with Emergency Diesel Generators and Off-Site supplies to the ESS buses unable to fulfill their intended function of allowing manual transfer between power supplies to the ESS buses. This resulted in the inability of Unit 1 and Unit 2 to comply with SR (Surveillance Requirements) 3.8.1.8 and 3.8.1.16 thus requiring declaration of Operating AC Sources inoperable. This condition could have prevented the fulfillment of the safety function of systems required to maintain the reactor in a safe shutdown condition, remove residual heat, control the release of radioactive material, and mitigate the consequences of an accident. Subsequent actions were taken in accordance with station procedures to remove fuses for the affected sync circuit, restoring the manual transfer function to all but the Unit 2 B ESS bus. One Emergency Diesel and one offsite source remain inoperable with the fuses removed. At no time were any ESS buses disconnected from offsite power. All ESS buses remained capable of being automatically energized from their respective emergency diesel in an emergency. The licensee has notified the NRC Resident Inspector. | |
| ENS 52327 | 28 October 2016 14:24:00 | Cooper | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Shutdown Cooling Decay Heat Removal | At 0851 (EDT) on October 28, 2016, Division 1 RHR was started in shutdown cooling (SDC) mode of operation. Prior to starting the RHR system, the Alternate Decay Heat Removal (ADHR) system was maintaining RPV and Spent Fuel Pool temperature. At 0924 on October 28, 2016, RHR (pump A) tripped due to RHR-MOV-17 (SDC suction valve) closing. This is considered to be an event or condition that could have prevented fulfillment of a safety function, and is reportable under 10 CFR 50.72(b)(3)(v)(B). RHR SDC subsystem A was declared inoperable. CNS (Cooper Nuclear Station) entered LCO 3.9.7, Condition A - Required Action A.1: Verify an alternate method of decay heat removal is available within 1 hour and once per 24 hours thereafter; Condition C - Required Action C.1: Verify reactor coolant circulation by an alternate method within 1 hour from discovery of no reactor coolant circulation and once per 12 hours thereafter, and Required Action C.2: Monitor reactor coolant temperature hourly. All LCO conditions specified have been met. ADHR remained in service throughout the event and the plant remained aligned for natural circulation. Spent fuel pool weir temperature monitoring was commenced to verify natural circulation. No increase in RPV (reactor pressure vessel) temperature has been observed. There was no impact to plant operations. Initial investigation indicates that installation of PCIS relay K27 during a maintenance activity physically agitated the adjacent relay, K30, which actuated and caused RHR-MOV-17 to close. The NRC Resident Inspectors have been informed. | |
| ENS 52271 | 30 September 2016 02:00:00 | Arkansas Nuclear | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Residual Heat Removal Decay Heat Removal | At 2100 CDT on 09/29/16, while in Mode 6, both trains of Decay Heat (Residual Heat Removal) were declared inoperable due to a cracked weld on a 1" common pipe. The leak developed in a USAS B31.7, Class1 pipe at a weld upstream of pressure indication isolation valve DH-1037. The leak is not isolable from the common 8-inch Decay Heat piping and encompasses approximately 1/3 (one third) of the pipe circumference. At the time of discovery, the unit was in Lowered Inventory with both Loops of Decay Heat in service. Subsequently, one train of Decay Heat has been secured to reduce the likelihood of crack propagation. One Train of Decay Heat remains in service providing the function of removing Decay Heat and the other train is readily available. The leakage impacts redundant equipment required to fulfill a safety function. In the current condition, both trains are required to be operable to meet Technical Specification LCO 3.9.5, Decay Heat Removal (DHR) and Coolant Circulation-Low water Level. This condition is reportable per 10 CFR 50.72(b)(3)(v)(B) for any event or condition that results in a loss of Safety Function associated with the Decay Heat System (Residual Heat Removal System). The licensee has notified the NRC Resident Inspector. The leak is approximately 0.25 gallons per minute and pipe pressure is 140 psi. Compensatory measures are in place and include an individual posted to watch the pipe in case plugging is necessary. Repairs to the pipe will be completed once pipe is able to be drained. | |
| ENS 52247 | 16 September 2016 20:57:00 | Davis Besse | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | At 1657 Eastern Daylight Time (EDT) the plant entered Mode 4 (from Mode 5), and subsequently, at 1710 EDT, it was discovered that 480V AC essential busses E1 and F1 were being supplied from the shutdown operations transformers. The essential busses E1 and F1 are required to be aligned to the power operations transformers in Mode 4 for operability in accordance with TS 3.8.9. With both E1 and F1 essential busses aligned to the shutdown operations transformers with the plant in Mode 4, both trains of the essential electrical power distribution system were inoperable, resulting in a loss of safety function. At 1733 EDT both E1 and F1 essential busses were aligned to the power operations transformers as required by TS 3.8.9. This issue is being reported as a loss of safety function of the essential electrical busses. The NRC Resident Inspector has been notified of the event.
Engineering reviewed the actual conditions during the approximate 36 minutes the 480V AC essential busses were being supplied from the shutdown operations transformers. Grid voltages were higher than assumed minimum voltages, and electrical loading during Mode 4 conditions were reduced from expected full power operation loading. As a result, Engineering determined that all equipment remained capable of performing its required functions while connected to the shutdown operations transformers. Because the equipment remained capable of satisfying the requirements for Operability, no condition existed that could have prevented the fulfillment of a safety function. Therefore, no loss of safety function existed for the 480V AC essential buses, and the notification made per 10 CFR 50.72(b)(3)(v)(A-D) by the Davis-Besse Nuclear Power Station on 9/16/2016 (EN# 52247) is being retracted. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. Notified the R3DO (Jeffers). | ||
| ENS 52242 | 15 September 2016 14:08:00 | Arkansas Nuclear | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(ii)(B), Unanalyzed Condition | Residual Heat Removal | During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornadoes, Arkansas Nuclear One, Unit 1, identified non-conforming conditions in the plant design such that specific TS equipment on Unit 1 is considered not (to) be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering vital switchgear rooms 99 and 100 and striking vital switchgears in the rooms. A tornado could generate multiple missiles capable of striking the Unit 1 vital switchgear and rendering both safety related AC electrical trains inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for 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 (A) safe shutdown capability, (B) residual heat removal capability, or (D) accident mitigation. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Licensee Event Report (LER) 50-313/2016-002-00 was recently submitted addressing previously identified tornado missile vulnerabilities at the Unit 1 plant. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector. | |
| ENS 52234 | 11 September 2016 19:51:00 | Arkansas Nuclear | 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)(3)(ii)(B), Unanalyzed Condition | Feedwater | During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Arkansas Nuclear One, Unit 1, identified non-conforming conditions in the plant design such that specific TS equipment on Unit 1 is considered not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Unit 1 Controlled Access area, elevation 386', Upper North Electrical Penetration Room (UNEPR) through penetrating a hollow metal door and then striking safety related cables. A tornado could generate multiple missiles capable of striking the Unit 1 UNEPR and rendering both safety related emergency feedwater trains inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(B) and (D) for 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 (B) Remove residual heat, or (D) Mitigate the consequences of an accident. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Licensee Event Report (LER) 50-313/2016-002-00 was recently submitted addressing previously identified tornado missile vulnerabilities at the Unit 1 plant. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector. A similar evaluation is on going for Unit 2. | |
| ENS 52195 | 24 August 2016 13:41:00 | Arkansas Nuclear | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(3)(ii)(B), Unanalyzed Condition | During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Arkansas Nuclear One, Unit 1, identified non-conforming conditions in the plant design such that specific TS equipment on Unit 1 is considered to not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Unit 1 Cable Spreading Room through penetrating the hollow metal door or potentially from spalling of the block wall separating Room 96 and 97. A tornado could generate multiple missiles capable of striking the Unit 1 Cable Spreading Room and rendering both safety related electrical trains inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for 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 (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Licensee Event Report (LER) 50-313/2016-002-00 was recently submitted addressing previously identified tornado missile vulnerabilities at the Unit 1 plant. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector. | ||
| ENS 52181 | 15 August 2016 19:52:00 | Pilgrim | 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 | Service water | On Monday, August 15, 2016 at 1552 (EDT), with the reactor at (about) 70 percent core thermal power (CTP), Pilgrim Nuclear Power Station (PNPS) entered a 24-hour shutdown Limiting Condition for Operation Action Statement (LCO-AS) for Salt Service Water (SSW) inlet temperature exceeding the Technical Specification (TS) limit in TS 3.5.B.4. The LCO-AS was subsequently exited at 1651 hours when the temperature of SSW trended to below the TS limit. Under certain design conditions, the SSW system is required to provide cooling water to various heat exchangers such as the Reactor Building Closed Cooling Water (RBCCW) and Turbine Building Closed Cooling Water (TBCCW) systems. When the inlet temperature to these supplied loads exceeds the 75 degrees F limit established in the TS, the SSW system is conservatively declared inoperable until the temperature trends below this value. This condition existed for approximately 60 minutes. The SSW temperature is being closely monitored and trended on a continuous basis. This event has no impact on the health and safety of the public. The licensee has notified the NRC Senior Resident Inspector. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(B) and 10 CFR 50.72(b)(3)(v)(D) due to an event or condition that could have prevented fulfillment of a safety function. The licensee will be notifying the Commonwealth of Massachusetts Emergency Management Agency. | |
| ENS 52133 | 29 July 2016 03:57:00 | Nine Mile Point | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | On 7/28/2016 at 2357 EDT, Nine Mile Point Nuclear Station Unit 1 experienced a fault in the in-service 11 RPS UPS (Reactor Protective System Ultimate Power Supply), resulting in an isolation of both emergency condensers. Emergency condenser 12 was returned to standby on 7/29/2016 at 0041 EDT and emergency condenser 11 was returned to standby on 7/29/2016 at 0045 EDT. This 8-hour non-emergency report is being made based upon requirements of 10CFR50.72(b)(3)(v)(B) which states, 'Licensee shall notify the NRC of 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: (B) Remove residual heat.' The state of New York and the NRC Senior Resident Inspector were informed. | ||
| ENS 51889 | 28 April 2016 00:07:00 | Millstone | 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 | Auxiliary Feedwater | On April 27, 2016 at 20:07 (EDT), a high energy line break (HELB) door between the turbine driven auxiliary feedwater pump and the motor driven auxiliary feedwater pump rooms was discovered open and unattended. In the event of a HELB, this condition could have rendered both trains of auxiliary feedwater inoperable. The boundary was not operable for approximately 1 hour. Upon discovery, the door was closed restoring the boundary. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v) as any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat; and (D) Mitigate the consequences of an accident. The NRC Senior Resident Inspector has been notified. The Licensee notified the Connecticut DEEP and Waterford Dispatch. | |
| ENS 51808 | 19 March 2016 22:10:00 | Arkansas Nuclear | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Two (2) potentially degraded flood barriers at penetrations 0073-01-0034 and 0073-01-0063 were identified in the area between the Unit 1 Turbine Building and Auxiliary Building. The deficient barriers are a 'blockout section' of the floor designed to house multiple penetrations that transition from the Turbine Building to the Auxiliary Building. Attempts have been made to investigate the status of the flood barrier with no definitive results. Investigations and additional evaluations are continuing, however, it is currently unknown if the aggregate of these two flood barriers could potentially overwhelm and flood the Auxiliary Building which would challenge equipment necessary to remove residual heat and constitute an unanalyzed condition. Based on current conditions (i.e., no forecast flooding conditions), this condition does not present an immediate safety concern. This condition has been determined to be reportable per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B). This condition is a non-emergency condition. This condition has been entered into the Corrective Action Program. Compensatory measures have been prepared to allow placement of a seal over the identified deficient barriers. If required these seals can be installed well in advance of forecast flood conditions. Permanent repairs are currently being designed for installation. The NRC Resident Inspector has been notified.
This is an 8 hour non-emergency supplemental notification to previously issued Event Notification number 51808. In EN 51808, two non-functional barriers were identified and reported, and during an extent of condition review, a third barrier has been identified that does not conform to expected flood barrier standards A potentially degraded flood barrier at 'blockout' penetration 0073-01-9018 was identified in the area between the Unit 1 Turbine Building and Auxiliary Building. The deficient barrier is a 'blockout section' of the floor designed to house multiple penetrations that transition from the Turbine Building to the Auxiliary Building. Attempts have been made to investigate the status of the flood barrier with no definitive results. Investigations and additional evaluations are continuing; however, it is currently unknown if this flood barrier could potentially be overwhelmed and flood the Auxiliary Building which would challenge equipment necessary to remove residual heat and constitute an unanalyzed condition. Based on current conditions (i.e., no forecast flooding conditions), this condition does not present an immediate safety concern. This condition has been determined to be reportable per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B). This condition is a non-emergency condition. This condition has been entered into the Corrective Action Program. Compensatory measures have been prepared to allow placement of a seal over the identified deficient barrier. If required this seal can be installed well in advance of forecast flood conditions. Permanent repairs are currently being designed for installation. The NRC Resident Inspector was notified earlier in the evening that this event would be updated. Notified the R4DO (Haire). | ||
| ENS 51807 | 18 March 2016 16:28:00 | Fort Calhoun | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Shutdown Cooling Containment Spray | During a scheduled surveillance test on 3/18/2016 at 1128 (CDT), Fort Calhoun ultrasonic testing technicians discovered a void on the common shutdown cooling heat exchanger discharge piping. This piping is normally isolated during power operation, and the void does not adversely affect the Containment Spray function, Low Pressure Safety Injection function, or High Pressure Safety Injection function. This isolated piping with the void is placed in service only during shutdown cooling operation. The fluid height measured was 10.8 inches, compared to the required height of 11.7 inches for the surveillance test. The void could potentially complicate the initiation of shutdown cooling in the required mode of operation. This piping was last tested satisfactory on 12/31/2015. The source of the void is still under investigation. Fort Calhoun maintenance was successful in venting the void on 3/18/2016 at 1704 CDT. The NRC Resident Inspector has been notified.
Following the 8-hour 10 CFR 50.72 notification made on 3/18/16 (EN 51807), further engineering analysis has determined that the ensuing water hammer transient would not have prevented the shutdown cooling system from performing its required safety functions. Specifically, it was found that the resulting system pressure transient would not cause any relief valves to lift and that piping and supports would not be significantly challenged. Therefore, the common shutdown cooling heat exchanger discharge piping remained operable by the detailed analysis. As such, the safety function was not lost and the event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(B). Notified the R4DO (G Miller). | |
| ENS 51800 | 17 March 2016 20:15:00 | Grand Gulf | 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 | Service water Shutdown Cooling High Pressure Core Spray Core Spray Residual Heat Removal Decay Heat Removal | 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 51601 | 13 December 2015 08:07:00 | Harris | 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 | This is a non-emergency eight hour notification for a condition that could have prevented fulfillment of a safety function in a system needed to remove decay heat and to mitigate the consequences of an accident. At approximately 0307 EST on December 13, 2015, the 'A' and 'B' trains of the essential services chilled water (ESCW) system were inoperable for a time period of 23 minutes. The 'B' train of the ESCW system was being tested at the time that the 'A' train of the ESCW system became inoperable. The 'B' train of the ESCW system was returned to operable status at approximately 0330 EST and remains operable. The ESCW system provides cooling support functions to remove decay heat and mitigate the consequences of an accident. There were no safety system actuations during this time period and the plant is stable. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. The 'A' ESCW pump tripped on low lube oil pressure.
Upon further review of the condition, it was determined that the 'A' train of the ESCW system was capable of performing its safety function throughout the time period previously reported, and the 'A' train of the ESCW system was operable. Therefore, there was no loss of safety function. Event notification 51601 is retracted. The NRC Resident Inspector has been informed of this notification retraction. Notified the R2DO (Guthrie). |