Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56997 | 28 February 2024 18:50:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Automatic Start of Emergency Diesel Generators | The following information was provided by the licensee via phone and email: At 1350 EST on 2/28/2024, with Calvert Cliffs Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 65 percent power, an actuation of the '1A' and '2A' emergency diesel generators' auto-start occurred due to an undervoltage condition on the number 11 and number 21 4kV buses which are fed from the number 11 13kV bus. The '1A' and '2A' emergency diesel generators automatically started as designed when the 4kV buses' undervoltage signals were received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the '1A' and '2A' emergency diesel generators. There was no impact on the health and safety of the public or plant personnel. 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 undervoltage condition was caused by the feeder breaker to the number 11 13 kV bus opening during electrical maintenance. | Emergency Diesel Generator | |
| ENS 56991 | 24 February 2024 20:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Steam Generator 22 Feed Pump Trip | The following information was provided by the licensee via email: At 1546 EST, with unit 2 at 100 percent power, the reactor was manually tripped due to the '22' steam generator feed pump tripping. The trip was uncomplicated with all systems responding normally post-trip. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Operations responded using emergency operation procedure EOP-0, Post Trip Immediate Actions and EOP-1, Uncomplicated Reactor Trip and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. ESFAS (engineered safety features actuation systems) actuation (auxiliary feedwater manual actuation) is reportable under 10 CFR 50.72(b)(3)(iv)(A) 8-hour report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 56856 | 16 November 2023 07:27:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 0227 EST on 11/16/23, Calvert Cliffs Unit 2 experienced an automatic trip from the reactor protection system (RPS) based on reactor trip bus undervoltage (UV). At that time, a loss of U-4000-22 (13 kV to 4 kV transformer) caused a loss of 22, 23, and 24 4 kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV. The loss of 22 and 23 4 kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4 kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4 hour report ESFAS (engineering safety features actuation system) actuation (2B DG start on UV) is reportable under 10 CFR 50.72(b)(3)(iv)(A) - 8 hour report AFW operation is reportable under 10 CFR 50.73(a)(2)(iv)(A) - 60 day report The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. There was no impact on Unit 1 operations. Unit 2 is stable in mode 3.
ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8 hour report Notified R1DO (Defrancisco). | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 56841 | 8 November 2023 11:45:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Actuation of Auxiliary Feedwater System (AFW) | The following information was provided by the licensee via phone and email: At 0645 EST, on November 8, 2023, with Unit 2 in Mode 3 at zero percent power, a manual actuation of the auxiliary feedwater system (AFW) occurred during a planned plant cooldown. The reason for the AFW manual-start was a trip of the 22 steam generator feed pump due to a high casing level. The 23 AFW motor driven pump was manually started in accordance with implementation of AOP-3G, Malfunction of Main Feedwater System to restore steam generator levels. There was no impact to Unit 1. 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 AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No other systems were affected. No other compensatory or mitigation strategies implemented. Plant cooldown was the only significant evolution in progress. No impact to other technical specifications or limiting conditions for operation. All systems functioned as required. The electric plant is being supplied by offsite power with all diesel generators available. No significant increase in plant risk. There was nothing unusual or not understood. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 56839 | 7 November 2023 21:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to NON-SAFETY Related Bus Under Voltage | The following information was provided by the licensee via email: At 1617 on 11/7/2023, Calvert Cliffs Unit 2 experienced an automatic trip from a Reactor Protection System (RPS) based on reactor trip bus under voltage (UV). At that time a loss of U-4000-22 caused a loss of 22, 23, and 24 4kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV condition. The loss of 22 and 23 4kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4-hour report. ESFAS actuation (2B DG start on UV) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. Site Senior NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 was unaffected. Estimation of duration of shutdown is 24 hours. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 56386 | 2 March 2023 18:35:00 | 10 CFR 26.719, FFD Reporting requirements | FITNESS-FOR-DUTY Report - Supervisor Failed FITNESS-FOR-DUTY Test | The following information was provided by the licensee via phone and email: A non-licensed supervisor tested positive in accordance with the FFD (fitness-for-duty) testing program. The individual's authorization for site access has been terminated. The NRC Senior Resident Inspector has been notified. | ||
| ENS 56355 | 6 February 2023 11:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Communications and Assessment Capabilities | The following information was provided by Constellation via email: On 02/06/2023 at 0416 EST, the Constellation Emergency Response Organization (ERO) Notification Database System uploaded data files into the Mass Notification System (Everbridge) which is used to notify ERO personnel when activated. At 0630, the individual reviewing the uploaded files discovered that the data files did not upload properly and that Everbridge may not notify all ERO individuals within the required 10 minutes of system initiation. Constellation resolved the issue by 0752. During the time period of 0416 to 0752, control room operators would have been unaware that the ERO notification was not successful. Therefore, this issue constitutes a loss of offsite communications capability and is reportable under 10 CFR 50.72(b)(3)(xiii), 'The licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' This loss of offsite communications capability affected all Constellation nuclear stations. There was no impact on the health and safety of the public or plant personnel. Each affected station NRC Resident Inspectors have been or will be notified. | ||
| ENS 56347 | 25 December 2022 11:37:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-DAY Optional Telephonic Notification of Invalid Actuation of Emergency Diesel Generator (EDG) | The following information was provided by the licensee via email: This 60-day optional telephone notification is being made in lieu of an LER (Licensee Event Report) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0637 EST on December 25, 2022, the 2B EDG inadvertently started and ran unloaded without a valid undervoltage or safety injection actuation signal. It was determined that this condition was caused by the failure of the emergency start button due to age-related degradation. The button is normally held depressed (closed) by the glass enclosure in standby. To start the EDG using the Emergency Start Button, the button is released (open) when the glass enclosure is broken, which sends a start signal to the EDG. During troubleshooting, the resistance across the button contacts was measured at zero volts DC, indicating the button had failed to an open state causing the EDG to start. The button fell apart when the glass enclosure was removed. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the emergency diesel generator. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 56108 | 14 September 2022 17:48:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Safety Injection Actuation Signal While Shutdown | The following information was provided by the licensee via email: With Unit 2 in Mode 5 and cooling down for a scheduled maintenance outage, Operations was bypassing safety injection and safety injection bypass sensor modules per procedure. Due to a human performance issue, a safety injection actuation signal was generated. All equipment operated as expected. Operations has reset the safety injection actuation signal and restored the equipment to its required condition. Unit 1 was at 100 percent power and unaffected by this issue. The NRC Resident Inspector will be notified. | ||
| ENS 55684 | 3 January 2022 17:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Loss of Electrical Load | The following information was provided by the licensee via email: At 1223 (EST) on 01/03/2022, Calvert Cliffs Unit 2 automatically tripped from 100 percent power due to loss of electrical load. The cause is under investigation. The site Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods inserted and decay heat is being removed via the condenser. The plant is in a normal shutdown electrical lineup. There was no impact on Unit 1. | ||
| ENS 55597 | 21 November 2021 15:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Automatic Auxiliary Feedwater Actuation | At 1046 EST on November 21, 2021, with Calvert Cliffs Nuclear Power Plant Unit 2 in Mode 1 at 100 percent power, the reactor was manually tripped due to lowering levels in both steam generators following a loss of the 21 and 22 steam generator feed pumps. An Auxiliary Feedwater System actuation occurred to restore steam generator water levels. The trip was not complicated, with all systems responding normally. Decay heat is being removed by the Auxiliary Feedwater System. Calvert Cliffs Nuclear Power Plant Unit 1 is unaffected and remains in Mode 1 at 100 percent power. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification. RPS actuation, per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the automatic actuation of the Auxiliary Feedwater System is being reported as an eight-hour, non-emergency notification, Specific System Actuation, per 10 CFR 50.72(b)(3)(vi)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 55147 | 22 March 2021 02:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Lowering Steam Generator Level | At 2216 EDT on 3/21/2021, Calvert Cliffs Unit 2 was manually tripped from 37 percent power due to lowering level in the 21 Steam Generator. All systems responded per design. Main Feedwater was secured and Auxiliary Feedwater was manually initiated. The Site Senior Resident has been notified. The cause of the lowering level in the 21 Steam Generator is under investigation. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 55134 | 12 March 2021 04:15:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Report | A licensed operator had a confirmed positive alcohol test during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspectors have been notified. | ||
| ENS 55122 | 4 March 2021 08:23:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Notification of Degraded Condition | At time 0323 (EST) on March 04, 2021, it was determined that the Reactor Coolant System (RCS) pressure boundary did not meet the acceptance criteria under ASME, Section XI IWB-3600, "Analytical Evaluation of Flaws." This condition will be resolved prior to plant start up. This event is being reported as an eight hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident has been notified. | Reactor Coolant System | |
| ENS 55020 | 3 December 2020 14:23:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of an Emergency Diesel Generator | At 0923 EST on December 3, 2020, with Unit 1 in Mode 1 at 100 percent power, an actuation of the Emergency AC Electrical Power System (Emergency Diesel Generator 1A) occurred during normal plant operations. The reason for Emergency Diesel Generator 1A auto start was due to Class 1E 4KV Bus 11 feeder breaker opening. The Emergency Diesel Generator 1A automatically started as designed when the loss of voltage signal on 4KV Bus 11 was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the Emergency AC Electrical Power System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The cause of the 4KV Bus 11 Feeder Breaker opening is unknown at the present time and is under investigation. | Emergency Diesel Generator | |
| ENS 54979 | 3 November 2020 05:11:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Both Trains of Salt Water Inoperable | At 0011 EST on 11/03/20, it was discovered that BOTH trains of salt water were simultaneously INOPERABLE. While in a planned (limiting condition for operation) LCO window with the 21 salt water train INOPERABLE for post-maintenance testing, debris intrusion in the 22 salt water header rendered the redundant salt water train 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). One train of salt water was restored to operable at time 0026 EST. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. This event did not affect Unit 1.
Following the eight hour 10 CFR 50.72 notification made on 11/03/2020 (EN 54979), further engineering analysis determined that 22 Saltwater subsystem flow remained at levels sufficient to fulfill its safety function based on the conditions existing at the time of the event. Despite flow in 22 Saltwater subsystem falling below the short term (four hour) minimum value for approximately 15 minutes, engineering analysis was able to determine the increased heat removal capacity associated with the lower bay temperatures was sufficient to offset the reduced heat removal capacity associated with the lower 22 Saltwater subsystem flow. This demonstrated that actual heat transfer to the saltwater subsystem was sufficient to ensure all safety functions were fulfilled during the event. Therefore, this event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(A), (B) and (D). The NRC Resident has been informed. Notified R1DO (Greives) | ||
| ENS 54325 | 11 October 2019 17:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Initiated | At 1300 EDT, a Technical Specification required shutdown was initiated at Calvert Cliffs Unit 1. Technical Specification Action 3.1.4.C (Restore Control Element Assembly (CEA) alignment) was entered on 10/11/2019 at 1100 EDT, with a Required Action to reduce thermal power to less than 70 percent Rated Thermal Power and restore CEA alignment within 2 hours. This Required Action was not completed within the Completion Time; therefore, a Technical Specification required shutdown was initiated, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). At 1345 EDT, CEA alignment was restored and Technical Specification 3.1.4 (Control Element Assembly Alignment) was met. Reactor Power is being stabilized. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | ||
| ENS 53921 | 9 March 2019 17:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Sewage Line Backup Causing Wastewater Discharge to Chesapeake Bay | A sewage line on the south end of the plant backed up causing sanitary wastewater to flow into storm drains and out to the Chesapeake Bay. This is a required notification of the Maryland Department of the Environment under COMAR (Code of Maryland Regulations) 26.08 for discharge of a pollutant into navigable waters or the adjoining shoreline. The amount has been estimated at less than 1000 gallons and the source has been isolated and storm drains have been covered to stop any flow into them and subsequently to the Chesapeake Bay. This notification is made in accordance with 10CFR50.72(b)(2)(xi) due to notification of a state agency. The licensee notified the NRC Resident Inspector. | ||
| ENS 53663 | 11 October 2018 04:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Air Start Solenoid Closure Failure | During a post maintenance start of the 1B diesel generator, the air start solenoid valves did not close as expected. This resulted in lowering air pressure in the common air start headers causing inoperability of the 2A and 2B diesel generators at time 23:03. The 1B diesel generator was isolated from the common air start header, which restored the air start header pressure to the 2A and 2B diesel generators. The 2A and 2B diesel generators were declared operable at 23:34. The NRC Resident Inspector was notified. | ||
| ENS 53493 | 7 July 2018 04:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Ventilation Inoperable | At 0242 EDT, the CCNPP (dual unit, single control room) control room supply damper failed shut. This rendered the Unit 1 and Unit 2 control room ventilation inoperable and the appropriate LCOs were entered. At 0249, control room ventilation was restored to service and the appropriate LCOs were exited. This is being reported under 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. | ||
| ENS 53401 | 14 May 2018 17:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Saltwater System Non-Conforming with Licensing Basis for Tornado Generated Missiles | On May 14, 2018, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of a tornado generated missile, Calvert Cliffs identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from a tornado generated missile. A tornado could generate a missile that could strike the Unit 1 Saltwater system header and associated piping. This could result in damage to the unit 1 Saltwater system header which could affect the ability of the Unit 1 Saltwater subsystems to perform their design function if such a tornado would occur. 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)(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 mitigate the consequences of an accident. This condition is being addressed in accordance with NRC enforcement guidance provided in EGM 15-002 and DSS-ISG-2016-01. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification. | ||
| ENS 52640 | 25 March 2017 05:42:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Refueling Water Tank Level Inadvertently Lowered Below Ts | While performing a purification subsystem alignment on the Unit-2 Refueling Water Tank, an inadvertent transfer of Refueling Water Tank level to the common Spent Fuel Pool occurred. This transfer resulted in lowering Unit-2 Refueling Water Tank level below the Technical Specification (TS) required limit for the current mode of operation at 0142 (EDT) on 3/25/17. Upon recognition of the inadvertent transfer, Operations secured the lineup and restored Unit-2 Refueling Water Tank level to its normal operating band at 0225 on 3/25/17. This event is reportable under 10 CFR 50.72(b)(3)(v)(D) '...any event or condition that at the time of discovery could have prevented the fulfillment of the safety function structures or systems that are needed to mitigate the consequences of an accident.' With less than the required Technical Specification volume in the Refueling Water Tank, insufficient volume existed in the Refueling Water Tank to maintain 30 minutes of full flow Safety Injection, and subsequent continued pump operation after transition to recirculation mode of operation. This level is required by Technical Specification 3.5.4.B and has a one hour action statement to restore level. The NRC Resident Inspector has been notified.
The purpose of this notification is to retract ENS notification 52640 made on March 25, 2017 for Calvert Cliffs. After further evaluation, it has been determined that the volume of water in the Unit 2 Refueling Water Tank was never below the TS required volume of 400,000 gallons. The evaluation considered the as-found condition of the level transmitter and the existing environmental conditions of the tank in determining the actual RWT water volume on the day of the event. Therefore, this event does not meet the criteria of 10 CFR 50.72(b)(3)(v)(D) and the ENS report is being retracted. The licensee will notify the NRC Resident Inspector. Notified R1DO (Cook). | ||
| ENS 52409 | 5 December 2016 11:38:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty | A non-licensed supervisory employee had a confirmed positive for a prohibited substance during a random fitness-for-duty test. The individual's unescorted access to the plant has been denied. The licensee has notified the NRC Resident Inspector. | ||
| ENS 52406 | 4 December 2016 03:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Caused by Main Turbine Trip | On 12/3/16 at 2224 EST, Calvert Cliffs Unit-2 experienced an automatic reactor trip from full power due to a leak in the Unit-2 Main Turbine Electro-Hydraulic Control (EHC) system. The EHC leak caused the Unit-2 Main Turbine governor valves to close, resulting in a turbine trip and automatic reactor trip. The site Outage Control Center is manned, and investigation into the cause of the leak is underway. Unit-2 remains stable in Mode 3 with normal heat removal. Unit-1 remains at full power and was not affected by the trip. The plant is in a normal shutdown electrical lineup. All Control rods fully inserted and no primary or secondary safety relief valves lifted during the trip. The licensee has notified the NRC Resident Inspector. The licensee will be notifying Calvert County. | Main Turbine Control Rod | |
| ENS 51967 | 31 May 2016 20:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Calvert Cliffs Loss of Load Turbine/Reactor Trip from 100 Percent Power | On 05/31/2016 at 1626 (EDT), a Reactor Trip occurred on Loss of Load RPS (Reactor Protection System) actuation. The Turbine Trip was apparently caused by a failed ESFAS (Engineered Safety Feature Actuation System) logic module. The failed NSR (Non Safety Related) logic module was associated with 11 Steam Generator High Level Turbine Trip. All systems responded as designed. Normal decay heat removal is to the condenser. All offsite power sources remained in service. The NRC Resident Inspector was notified. | Steam Generator Decay Heat Removal | |
| ENS 51747 | 20 February 2016 08:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Discovered on Pressurizer Safety Relief Valve Weld | During the CCNPP (Calvert Cliff Nuclear Power Plant) Unit 1 Refueling, (an ultrasonic test) examination of the Unit 1 dissimilar metal weld (DMW) of the pressurizer safety relief valve RV-201 safe end weld found an indication that is approximately 80 percent though-wall, inside diameter connected axial flaw in weld #4-SR-1006-1. This is an ASME Class 1 component and Unit 1 has entered the requirements of Technical Requirements Manual (TRM) 15.4.3.A. CCNPP Event #4254. The licensee has notified the NRC Resident Inspector. | Safety Relief Valve | 05000317/LER-2016-002 |
| ENS 51683 | 25 January 2016 08:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Elevated Condenser Sodium Levels | At 0315 EST on 1/25/16, Calvert Cliffs Unit 1 was manually tripped from 10 percent power due to elevated condenser sodium levels. All systems responded per design. Main Feed was secured and auxiliary feed water was initiated. The elevated sodium levels are believed to be due to a condenser tube leak. The reactor is currently shutdown and stable in Mode 3 and will remain in Mode 3 until repairs are effected. Unit 2 was not affected and remains at full power. The Licensee has notified the NRC Resident Inspector. | Auxiliary Feedwater | |
| ENS 51577 | 1 December 2015 23:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip After Steam Generator Feed Pump Trip | On 12/01/2015 at 1820 EST, the Main Control Room received a 22 Steam Generator Feed Pump trip. The 22 Steam Generator Feed Pump was not able to be reset and the Main Control Room manually tripped the Unit 2 Reactor. The licensee entered Emergency Operating Procedure (EOP)-0, 'Post Trip Immediate Actions' and all safety functions were met. At 1833, Unit 2 transitioned into EOP-1, 'Uncomplicated Reactor Trip.' At 1841, Unit 2 transitioned into Operating Procedure #4 , 'Plant Shutdown from Power to Hot Stand-by.' The plant is stable in Mode 3. All control rods inserted fully on the reactor trip. No primary or secondary safety relief valves lifted. The steam generators are being fed by the 21 steam generator feed pump and decay heat is being dumped to the condenser via the steam dumps. The electric plant is in a normal shutdown electrical lineup and there was no impact on Unit 1. Unit 1 continues to operate at 100 percent power. The cause of the 22 steam generator feed pump trip is still under investigation. The licensee notified the NRC Resident Inspector. | Steam Generator Safety Relief Valve Control Rod | |
| ENS 51510 | 2 November 2015 14:20:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Radiation Monitoring Equipment Due to Planned Maintenance | A planned modification at Calvert Cliffs Nuclear Power Plant will remove the Spent Fuel Pool Area Radiation Monitoring system from service. The planned out of service window is expected to be 7 weeks, beginning 11/02/15 and ending 12/18/2015. During this time portable radiation monitors will be installed as compensatory measures to monitor and support timely and accurate EAL declarations. Affected Emergency Response Organization personnel have been made aware of the compensatory measures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to a planned event that affects Emergency Preparedness Assessment Capability for greater than 72 hours. An update will be provided once the Spent Fuel Pool Area Radiation Monitoring system is restored to normal operation. The licensee has notified the NRC Resident Inspector.
This is concerning the planned modification to Calvert Cliffs Nuclear Power Plant Spent Fuel Area Radiation Monitoring System. The modification window is extended to January 8, 2016. Compensatory measures that were put in place on November 2, 2015 will continue to be in effect until the completion of the modification testing and the new equipment is fully in service. The licensee notified the NRC Resident Inspector. Notified R1DO (Ferdas).
This is concerning the planned modification to Calvert Cliffs Nuclear Power Plant Spent Fuel Area Radiation Monitoring System. The modification window was originally extended to January 8, 2016 as updated on December 18, 2015. Based on field equipment issues during the installation process, the modification window has been extended to February 12, 2016. Compensatory measures that were put in place on November 2, 2015 will continue to be in effect until the completion of the modification testing and the new equipment is fully in service. The licensee notified the NRC Resident Inspector. Notified R1DO (Dimitriadis). | ||
| ENS 50961 | 7 April 2015 16:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 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 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Dual Unit Automatic Reactor Trips Due to a Voltage Transient Resulting in Generator Trips | A loss of Main Generator Load which caused a Reactor Trip on Units 1 & 2. A switchyard voltage transient from a highline occurred, which caused an undervoltage condition on both units' safety related 4KV buses. Unit 1 is on normal heat removal to the condenser. Unit 2 is on auxiliary feedwater and normal condenser bypass valves for temperature control. An Auxiliary Feedwater Actuation System (AFAS) actuation occurred on Unit 2. The (Unit 2) 2B emergency diesel generator did not start and load on its respective 24-4 KV bus. The 24-4KV Bus was repowered from the alternate feeder breaker. Cause of the emergency diesel failure to start is under investigation. All safety functions are met for both units. All control rods fully inserted. The site is in a normal shutdown electrical configuration powered from offsite. The site plans to stay in Mode 3 pending restart. The licensee notified the NRC Resident Inspector, State and local authorities. A press release is planned.
During post trip review, it was determined that the 21 saltwater pump had to be manually started. With the failure of 2B emergency diesel generator, there were no saltwater pumps running for approximately 12 minutes. Additional troubleshooting determined the 2A emergency diesel generator sequencer did not automatically start 21 saltwater pump. The 2B emergency diesel generator was returned to service on 4/8/2015 at 1730 (EDT). The loss of saltwater (pump) and emergency diesel generator is reportable as an event that could have prevented fulfillment of a safety function and is also an unanalyzed condition. The licensee has notified the NRC Resident Inspector. Notified R1DO (Ferdas), IRD MOC (Grant), NRR EO (Morris). | Emergency Diesel Generator Auxiliary Feedwater Control Rod | 05000317/LER-2015-002 |
| ENS 50752 | 22 January 2015 22:45:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Heat Exchanger Lineup Could Exceed Design Basis Temperatures | On 1/20/15, it was determined that a certain line up of component cooling heat exchangers and shutdown cooling heat exchangers could exceed the design basis temperatures for the component cooling water system following a design basis accident. Although not a safety concern at this time because of low ultimate heat sink temperatures (which cools component cooling water), in the past the ultimate heat sink temperatures have been high enough to create this condition. This particular heat exchanger line up was unanalyzed in that the ultimate heat sink temperature limits were not known until 1/22/15. This issue has been entered into the corrective action program. A review of Control Room logs for 2014 showed that in 1 instance for Unit 1 and 1 instance for Unit 2, the Units were in an unanalyzed lineup with ultimate heat sink temperature greater than the maximum now calculated. During these instances, both Units had an unanalyzed condition that had potential to significantly degrade plant safety and is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
Further engineering analysis has refined the ultimate heat sink temperature that provides an acceptable safety system response with the component cooling water and shutdown cooling heat exchanger lineups in question. The revised information demonstrates that the system lineups that occurred in the last 12 months did not result in an unanalyzed condition that significantly degrades plant safety. This event notification is retracted. The NRC Resident Inspector will be notified. The R1DO (Burritt) was notified.
The retraction statement provided on 3/3/15 incorrectly addressed system lineups limited to the last 12 months. The original withdrawal was prematurely submitted and therefore the original notification (Event report 50752) is not retracted and remains valid. The NRC Resident Inspector has been notified. The R1DO (Cook) was notified. | Shutdown Cooling | |
| ENS 50687 | 17 December 2014 09:17:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Main Plant Vent Radiation Monitoring Unavailable | At 0417 EST on December 17, 2014, the Shift Manager was notified by Radiation Protection (RP) personnel that the Unit 2 AMP-200 radiation monitor used for backup monitoring of the Unit 2 Main Plant Vent when the Unit 2 Wide Range Noble Gas Monitor (WRNGM) is (Out of Service) OOS had experienced a malfunction that caused it to spike high for approximately five minutes before returning to a baseline value of 0 mR/hr. At the same time, the Unit 2 WRNGM was already out of service for the completion of a planned surveillance test. The Unit 2 WRNGM had been removed from service at 0740 EST on December 16, 2014. While the WRNGM is OOS, the AMP-200 radiation monitor is procedurally relied upon for being used as the means for determining Alert, Site Area Emergency, and General Emergency entry criteria under the applicable Emergency Action Level (EAL) Initial Conditions (ICs). The Unit 2 Main Vent Gaseous Monitor is procedurally relied upon for being used as the means for determining Unusual Event entry criteria for the same EAL ICs and remained in service during this time. After investigation by site RP and Maintenance personnel, the functionality of the Unit 2 AMP-200 was restored at 0700 EST on December 17, 2014 following a successful source check of the instrument. The investigation identified that the AMP-200 cable contacts were found wet and were subsequently cleaned and dried. Actions have been taken to prevent reoccurrence of the wetting condition. During the time from 0417 to 0700 EST on December 17, 2014, Unit 2 had a condition that represented a loss of emergency assessment capability and is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified and the NRC the notification time with the NRC Headquarters was 1210 EST. | ||
| ENS 50502 | 1 October 2014 05:24:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Partial Loss of Communications in Emergency Operations Facility and Joint Information Center | At 0750 EDT on October 1, 2014, the Shift Manager was notified that site Information Technology (IT) personnel were being mobilized to investigate a potential voice and network loss at the Emergency Operations Facility (EOF) and Joint Information Center (JIC). Site IT personnel were notified by offsite IT resources at 0727 EDT on October 1, 2014 of the issue that was first identified by IT monitoring software at 0124 on October 1, 2014. The site IT personnel that responded to the EOF and JIC reported to site Control Room and Emergency Preparedness (EP) personnel at 0845 that connectivity to the Exelon network and the internet was unavailable at both the EOF and the JIC. This loss of connectivity would prevent the ability of the EOF Emergency Response Organization (ERO) personnel to directly monitor key plant parameters via the site's Plant Process Computer (including the Site Parameter Display System) and other network-based plant parameter display systems. Site IT and EP personnel determined that the following communications equipment was not impacted by the connectivity issue: - Dedicated Offsite Agency Phones (primary method for contacting state and local agencies) - Commercial Phones and dedicated bridge line (primary method for contacting other site Emergency Response Facilities) - FTS-2001 Phones (e.g., ENS and HPN lines) - ERDS Additionally, EP personnel verified with Dose Assessment Office personnel that dose assessment and dose monitoring functions from the EOF could still be performed without delay. Site IT personnel reported to the Control Room at 1135 that connectivity to the Exelon network and the internet had been restored to a fully functional status. While site and fleet IT personnel continue to address and verify all appropriate corrective actions have been taken to prevent recurrence of the connectivity issue, the site has employed appropriate compensatory measures to ensure that the verbal transmission of key plant parameters from the site (Technical Support Center or Control Room) to the EOF is recognized and maintained. The NRC Resident Inspector has been notified.
This update retracts Event Report #50502, which reported that a loss of connectivity to the Exelon network and internet at the Emergency Operations Facility (EOF) and Joint Information Center (JIC) had impacted the ability of staff in these facilities to directly monitor key plant parameters via the site's Plant Process Computer and other network-based plant parameter display systems. Subsequent to the identification of this event, further investigation by site and fleet staff determined that adequate direction was included in applicable Emergency Response Organization (ERO) procedures to respond to data display system failures of this type. Specifically, the checklist (procedure) for the Operations Communicator in the EOF provided adequate direction for this ERO member to obtain required plant data from the Operations Communicator located in the Control Room via alternate methods (e.g., over the phone - phone lines remained functional throughout the time that the loss of computer connectivity condition existed). Therefore, this event did not result in a major loss of emergency assessment capability and was not reportable to the NRC under 10CFR50.72(b)(3)(xiii). The NRC Resident Inspector has been notified. Notified R1DO (Bickett) and Cyber Assessment Team via email. | ||
| ENS 50307 | 25 July 2014 03:15:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification Required Shutdown Due to Reactor Coolant System Pressure Boundary Leakage | At 2315 EDT on July 24th, 2014, CCNPP (Calvert Cliffs Nuclear Power Plant) U1 identified RCS (Reactor Coolant System) pressure boundary leakage from the instrument line to 1-PDT-123A, 11A reactor coolant pump differential pressure transmitter. Technical Specification 3.4.13, Action B was entered and requires that the unit be placed in Mode 3 within 6 hours and Mode 5 within 36 hours. (CCNPP has) initiated plant shutdown in accordance with this Technical Specification. Therefore, this is reportable under 10 CFR 50.72(b)(2)(i) Plant Shutdown Required by Technical Specifications. This is also reportable under 50.72(B)(3)(ii)(A) as a material defect in the primary coolant system that cannot be found acceptable under ASME Section XI, IWB-3600 or ASME Section XI, Table IWB-3410-1. The licensee notified the NRC Resident Inspector.
The licensee reduced reactor power to 10 percent and entered containment. The pressure boundary leak was isolated and the licensee exited the Technical Specification 3.4.13, Action B. The Technical Specification required shutdown was terminated. The licensee is making preparation to increase power to full load. The licensee will notify the NRC Resident Inspector. Notified the R1DO (Gray). | Reactor Coolant System | |
| ENS 50078 | 1 May 2014 14:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 1016 (EDT) CCNPP (Calvert Cliffs Nuclear Power Plant) Unit 1 automatically tripped due to an RPS actuation. Cause is under investigation. All safety functions are met with normal heat removal. Electric plant is in a normal lineup. No ESFAS (Engineered Safety Feature Actuation System) actuations have occurred. Steam Generator atmospheric dump valves momentarily opened and then closed. There is no known steam generator tube leakage. All control rods fully inserted on the trip. There was no impact on Unit 2 from this event. The licensee notified the NRC Resident Inspector and provided a courtesy notification to the Calvert County Control Center. | Steam Generator Control Rod | |
| ENS 49889 | 10 March 2014 20:53:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | An extent of condition review identified that there were unfused 250 VDC circuits for the battery charger ammeters and the control circuits for various DC lube oil pumps which left the turbine building and went through the respective cable spreading room and up into the Control Room. These circuits are protected by 200 amp fuses and in some cases 100 amp fuses. The concern is that under the fire safe shutdown rules, it is postulated that a fire in one fire area can damage these circuits and cause short circuits without protection that would overheat the cables and possibly result in secondary fires in other fire areas where the cables are routed. The secondary fires could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. This condition is reportable as an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector. | 05000336/LER-2014-002 05000423/LER-2014-002 | |
| ENS 49754 | 22 January 2014 02:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Dual Unit Trip on the Loss of the "21" 13Kv Bus | Dual Unit Trip due to loss of '21' 13 KV bus . All safety functions are met for both units. Unit 1 remained with normal heat removal. Unit 2 lost power to its normal heat sink and is stable on Auxiliary Feed water and Atmospheric Dump Valves for temperature control. Both trips were automatic trips. Due to loss of power a Under Voltage actuation occurred on both units ('14' and '24' 4Kv bus). Due to loss of main feed on Unit 2 a Auxiliary Feed water Actuation System (AFW) actuation occurred on Unit 2. Cause is under investigation. All control rods fully inserted on the loss of power to the Control Rod Drive Mechanisms (CRDMs). Both Units Reactor Coolant Pumps (RCPs) remained running during the transient. The normal Unit 2 heat sink was unavailable due to the loss of the operating circulating water pumps resulting in a loss of condenser vacuum. The Unit 2 AFW actuation included one of two steam-driven pumps and the motor-driven pump. Both Units Emergency Diesel Generators started and loaded and have since been secured. Both Units are stable and will remain in mode 3 (Hot Standby) pending the results of the investigation. The licensee will inform the NRC Resident Inspector. | Emergency Diesel Generator Auxiliary Feedwater Control Rod | 05000317/LER-2014-001 05000318/LER-2014-001 |
| ENS 49463 | 22 October 2013 14:59:00 | 10 CFR 26.719, FFD Reporting requirements | Random Fitness for Duty Confirmed Positive for a Non-Licensed Contract Supervisor | A non-licensed contract employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The contractor's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. | ||
| ENS 49449 | 18 October 2013 12:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | At 0830 EDT, Calvert Cliffs Nuclear Power Plant determined that the following was an unanalyzed condition: As a result of recent industry operating experience (OE 305419) regarding the impact of unfused direct current (DC) ammeter circuits in the Control Room, Calvert Cliffs performed a review of ammeter circuitry. The review determined the described condition to be applicable to Calvert Cliffs resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The wiring design for the ammeters contains a shunt in the current flow from each DC battery or charger. The ammeter wiring attached to the shunt does not contain fuses. It is postulated that a fire in the Unit 1 Cable Spreading Room could cause one of the ammeter wires to hot short to ground. Concurrently, the fire causes another DC wire from the opposite polarity on the same battery to also short to ground. This would cause a ground loop through the unfused ammeter cable. The potential exists that the cable could heat-up causing a secondary fire in the Unit 2 Cable Spreading Room. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to safely shutdown per 10 CFR 50 Appendix R. This condition is reportable as an eight hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified. | ||
| ENS 49424 | 10 October 2013 09:38:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Single Siren Activation | At 0538 (EDT on 10/10/13), Calvert Cliffs was notified by Calvert Control Center, that a siren in the Calvert County Area was activated. (It was) determined to be siren C-15 on Lloyd Bowen Road. This is 1 of 73 sirens. Per (procedure) CNG-NL-1.01-1004 this is a verbal report to Calvert Control Center. This event is reportable as a 4 hour non emergency notification 10CFR50.72(b)(2)(xi) as: any event resulting in notification to other government agencies that has or will be made. Heavy rains in the area are believed to have caused the siren activation. Maintenance personal have been dispatched to repair the siren. The licensee has notified the NRC Resident Inspector. | ||
| ENS 49323 | 5 September 2013 17:14:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Failure to Align Control Element Assembly During Testing | CCNPP (Calvert Cliffs Nuclear Power Plant) U2 CEA (Control Element Assembly) #27 fully inserted (CEA dropped) into the core during testing per STP 0-029-2. Technical Specification 3.1.4, Action B was entered and required the CEA to be realigned within 2 hours. With this action not met Technical Specification 3.1.4, Action C required the Unit be reduced to < 70% rated thermal power (achieved at 1308 (EDT)) and the CEA to be realigned within 2 hours. With the CEA being unable to be realigned, Technical Specification 3.1.4.F was entered at 1314 (EDT) which requires the Unit to be placed in Mode 3 within 6 hours. A plant shutdown has been initiated in accordance with this Technical Specification. Therefore, this is reportable under 10 CFR 50.72(b)(2)(i), 'Plant Shutdown Required by Technical Specifications.' The licensee will notify the NRC Resident Inspector. | ||
| ENS 49262 | 10 August 2013 11:11:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Inadvertent Single Emergency Siren Activation | Calvert Cliffs was notified at 0711 (EDT) by Calvert Control Center that a siren in the Calvert County Area was activated. (It was) determined to be Siren C15 on Lloyd Bowen Road. Per CNG-NL-1.01-1004 this is a verbal report to Calvert Control Center. This event is reportable as a 4-hour non-emergency notification (under) 10CFR50.72(b)(2)(xi) as 'Any event resulting in notification to other government agencies that has or will be made.' The licensee informed the NRC Resident Inspector. | ||
| ENS 49054 | 21 May 2013 09:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip After a Feed Pump Trip | Reactor trip. All safety functions met with normal heat removal. 22 SGFP (Steam Generator Feedpump) exhibited high vibrations and signs of coupling damage. Further investigation will be performed. All control rods fully inserted on the trip. Steam Generator level is being maintained with the remaining feedpump. Decay heat is being dumped to the main condenser. Electrical power is in the normal shutdown lineup. No relief or safety valves lifted during the trip. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector. | Steam Generator Main Condenser Control Rod | 05000318/LER-2013-004 |
| ENS 49012 | 9 May 2013 01:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The reactor automatically tripped at 2147 EDT. All control rods fully inserted on the trip and all systems responded as expected. Decay heat removal is to the main condenser. The plant is in its normal shutdown electrical lineup. The licensee is investigating the cause of the reactor trip. The licensee notified the NRC Resident Inspector. | Decay Heat Removal Main Condenser Control Rod | 05000318/LER-2013-003 |
| ENS 48790 | 27 February 2013 20:38:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Maintenance on Unit 2 Plant Process Computer | Calvert Cliffs Nuclear Power Plant will perform planned maintenance on the Unit 2 Plant Process Computer (PPC) and associated network infrastructure starting on February 27, 2013. The maintenance will remove the Unit 2 PPC for a planned duration of 5 days and will render the Unit 2 ERDS out of service for this timeframe. The current outage schedule has power being restored on 03/04/2013. Should an emergency be declared during this period, the Control Room will continue to have the capability to retrieve plant data inputs to assess plant conditions and perform core damage assessment at all times. Control Room Emergency Response personnel will use emergency response procedures to disseminate plant parameter data points to the effected Emergency Response Facilities until the U-2 PPC is restored. MIDAS (Meteorological Data) transmission will remain functional during the maintenance window. The Unit 2 ERDS data will not be available during the power outage from 02/27/2013 to 03/04/2013. All work associated with this plant data network software installation will be performed in an expeditious manner consistent with the goal of minimizing unavailability of the systems listed above. ERDS was functionally tested before the outage and will be in service as soon as power is restored. The licensee will notify the NRC Resident Inspector. | ||
| ENS 48763 | 17 February 2013 18:45:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leakage Discovered During Containment Walkdown | CCNPP (Calvert Cliffs Nuclear Power Plant) U-2 discovered during the Mode 3 NOP/NOT containment walkdown that 2CV-100F Pressurizer Spray valve had a pin hole leak on the leak-off welded cap. This indicates that the bellows has a leak and in addition, the packing is leaking, pressurizing the leak-off line thus the leak-off line is the pressure boundary making this pressure boundary leakage. Technical Specification 3.4.13 (Action B) requires the unit be in mode 3 in 6 hrs. Unit was in mode 3 when discovered and unit will be placed in mode 5 within 36 hrs. Currently, the plant is at 400 degrees and 950 PSIA. The licensee has notified the NRC Resident Inspector. | 05000318/LER-2013-001 | |
| ENS 48720 | 6 February 2013 14:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Maintenance on Plant Process Computer | This report is being made in accordance with 10CFR50.72(b)(3)(xiii). Calvert Cliffs Nuclear Power Plant will perform planned maintenance on the Unit 2 Plant Process Computer (PPC) and associated network infrastructure on February 06, 2013. The maintenance will remove the Unit 2 PPC for a planned duration of 12 hours and will render the Unit 2 SPDS out of service for this timeframe. Once the maintenance starts, the effected equipment and functions, including SPDS, can be returned to service within five minutes, if required. Should an emergency be declared during this period, the control room will continue to have the capability to retrieve plant data inputs to assess plant conditions and perform core damage assessment at all times. Control room emergency response personnel will use emergency response procedures to disseminate plant parameter data points to the affected Emergency Response Facilities until the U-2 PPC is restored. MIDAS (Meteorological Data) and Unit 2 ERDS (Emergency Response Data System) data transmission will remain functional during the maintenance window. All work associated with this plant data network software installation will be performed in an expeditious manner consistent with the goal of minimizing unavailability of the systems listed above. Back-out criteria has been identified as part of the work package. A test of all systems will be performed at the completion of the upgrade. The licensee notified the NRC Resident Inspector.
The plant process computer was returned to service at 1700 EST. The licensee will notify the NRC Resident Inspector. Notified the R1DO (Powell). | ||
| ENS 48603 | 18 December 2012 19:36:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Partial Loss of Emergency Assessment During Planned Modifications | Calvert Cliffs Nuclear Power Plant will begin a planned modification of the Unit 1 and Unit 2 Plant Process Computers (PPC's) and associated network infrastructure on December 18, 2012. This includes the Emergency Response Data System (ERDS) communication with the NRC Operations Center. This work will require the TSC, OSC and subsequently EOF to lose data flow from the plant data network for a period of approximately 36 hours. During this time, the following systems will be impacted for the duration of maintenance: 1. ERDS will be out of service. 2. Plant Data will not reach the TSC and OSC from the plant data network. 3. TSC Computer data transfer to the EOF only will be out of service. 4. Chemistry DAS remote monitoring capability will be lost. 5. Plant parameter data will not be available on the site's LAN network. Once the schedule maintenance starts, the systems (listed above) can be returned to service within 1 hour. Should an emergency be declared during this period, the Control Room will continue to have the capability to retrieve plant data inputs to assess plant conditions and perform core damage assessment. Control Room Emergency response personnel will use emergency response procedures (ERPIP 106) to disseminate plant parameter data to the effected Emergency Response Facilities. MIDAS (Meteorological Data) will continue to be operational at the site. Compensatory measures exist within the Calvert Cliffs Emergency Response procedures to provide plant data in the event of an actual Emergency to the NRC Operations Center until the ERDS can be returned to service. The licensee notified the NRC Resident Inspector.
On 12/20/2012 at 0200 EST, the plant process computer and associated network infrastructure was restored to service. The licensee will notify the NRC Resident Inspector. Notified the R1DO (Hunegs). | Emergency Response Data System | |
| ENS 48446 | 25 October 2012 14:31:00 | 10 CFR 26.719, FFD Reporting requirements | Confirmed Positive Fitness for Duty | A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The NRC Resident Inspector has been notified. | ||
| ENS 48368 | 3 October 2012 12:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Outage Due to Planned Maintenance | This report is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the planned loss of an Emergency Response Facility (ERF). Calvert Cliffs will be implementing a scheduled modification to renovate the site's common Technical Support Center (TSC) which will temporarily remove it from service. The planned maintenance window is expected to be three weeks beginning on 10/3/12 and concluding on 10/26/12. Should an emergency be declared requiring TSC activation during this period, the procedurally identified alternate TSC will be staffed and activated (CCNPP Simulator). All TSC responders and other key ERO personnel have been briefed on this activity and the applicable contingency actions that will be required to be taken in their specific emergency response procedures. This condition has no adverse impact to the health and safety of the public. An update will be provided once the TSC has been restored to normal operation. The licensee has notified the NRC Resident Inspector.
The TSC has been restored to normal operation. The licensee will notify the NRC Resident Inspector. Notified R1DO (Caruso). |