Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56698 | 25 August 2023 21:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Assessment Capability | The following information was provided by the licensee via email: At approximately 1600 CDT on 8/25/2023, a partial loss of the commercial phone communications system occurred that affects the emergency notification system (ENS) and the functionality of an emergency response facility. This is an eight-hour, non-emergency notification of a loss of emergency assessment capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii). Communications via alternate methods were subsequently established. The telecommunications provider has not provided an estimated repair time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | ||
| ENS 56576 | 14 June 2023 14:26:00 | 10 CFR 26.719, FFD Reporting requirements | FITNESS-FOR-DUTY Report | The following is a summary of information provided by the licensee via email: A non-licensed, non-supervisory employee was identified bringing a prohibited item into the protected area. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. | ||
| ENS 56354 | 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 55857 | 23 April 2022 13:54:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Ultrasonic Examination Results - Reactor Vessel Head Penetration | The following information was provided by the licensee via email: At 0854 (CDT) on April 23, 2022, while performing volumetric inspections required by ASME Code Case N-729-6, a rejectable indication on Reactor Vessel Head Penetration 75 Core Exit Thermocouple (CETC) was identified. The indication is located inboard of the J-groove weld and is OD-initiated (outer diameter - initiated). 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 Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The repair is scheduled during the refueling outage.
The following information was provided by the licensee via telephone and email: The rejectable indication on Reactor Vessel Head Penetration 75 Core Exit Thermocouple (CETC) initiated on the outside diameter (OD) of the nozzle in an area that was not surface stress mitigated (peened). The indication was found to be acceptable for continued operation under CFR and ASME requirements and will not be repaired during this outage. The licensee notified the NRC Resident Inspector. Notified R3DO (Ziolkowski). | ||
| ENS 55769 | 6 March 2022 03:15:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Degradation of Technical Support Center | The following information was provided by the licensee: At 2115 CST on March 5, 2022 Byron Station Technical Support Center (TSC) emergency ventilation system supply fan belt failed. This failure affected the ability of the TSC ventilation system to maintain adequate radiological habitability in the event of an emergency with an airborne radiological release. All other capabilities of the TSC are unaffected by this condition. If an emergency was declared requiring TSC activation during this period, the TSC would be staffed and activated using existing emergency planning procedures. If the TSC becomes uninhabitable, the Station Emergency Director would relocate the TSC staff to an alternate TSC location in accordance with applicable procedures. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the discovered condition affected the functionality of an emergency response facility. The licensee notified the NRC resident inspector. | ||
| ENS 54710 | 13 May 2020 15:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Response Facility | At 1000 CDT on May 13, 2020, the Byron Station Technical Support Center (TSC) emergency ventilation system inlet isolation damper would not open as required to support system operation. This failure affected the ability of the TSC ventilation system to maintain adequate radiological habitability in the event of an emergency with an airborne radiological release. All other capabilities of the TSC were unaffected by this condition. If an emergency was declared requiring TSC activation during this period, the TSC would be staffed and activated using existing emergency planning procedures. If the TSC became uninhabitable, the Station Emergency Director would relocate the TSC staff to an alternate TSC location in accordance with applicable procedures. The TSC emergency ventilation system inlet isolation damper has been repaired and is now functional. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the discovered condition affected the functionality of an emergency response facility. The NRC Resident Inspector has been notified. | ||
| ENS 54020 | 22 April 2019 18:24:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Initiation of the Auxiliary Feedwater System in Response to a Loss of Feedwater | At 1324 CDT, on 4/22/2019, with unit 2 in Mode 3 at 0 percent power, an intentional manual initiation of the Auxiliary Feedwater System occurred in response to a loss of feedwater condition. The loss of feedwater condition occurred after the non-safety related Startup Feedwater Pump was secured due to high bearing temperatures. The A Train Auxiliary Feedwater Pump was started per procedure. The Auxiliary Feedwater System started and operated as designed following intentional manual initiation. 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 Auxiliary Feedwater System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Feedwater Auxiliary Feedwater | |
| ENS 53551 | 16 August 2018 05:00:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness-For-Duty Test Positive for Licensed Employee | In accordance with 10 CFR 26.719(b)(2)(ii), this notification reports a licensed Operations employee had a confirmed positive for alcohol during a random fitness for duty test. The individual was not in the protected area and not performing licensed duties at the time of discovery. The employee's access to the plant has been suspended. The NRC Resident Inspector has been notified. | ||
| ENS 53491 | 6 July 2018 05:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Transformer Failure Causes Loss of Offsite Power to Unit 2 | At 1201 (CDT), Station Auxiliary Transformer 242-2 experienced a bushing failure, resulting in a loss of offsite power to Unit 2. The 2A and 2B Diesel Generators started and sequenced loads onto the Unit 2 ESF buses appropriately. All other buses normally powered from the Station Auxiliary Transformers automatically transferred to the Unit Auxiliary Transformers. ESF Bus 241 and 242 Undervoltage Relays actuated to start the Diesel Generators and the 2A Auxiliary Feedwater Pump started on the 2A Diesel Generator sequencer. ESF Battery Charger 212 tripped at the same time, which was an unexpected condition. DC Bus 212 was cross-tied with DC Bus 112. This notification is being made under 10 CFR 50.72(b)3(iv)(A) due to the actuation of both Unit 2 Diesel Generators and the 2A Auxiliary Feedwater Pump. The NRC Senior Resident Inspector has been notified. Currently, offsite power was restored via the Unit 1 Unit Auxiliary Transformer. Both Unit 2 Emergency Diesel Generators have been secured. DC Busses are still cross-tied. The licensee is currently in a 72-hour shutdown action statement for the loss of offsite power and a 7-day action statement for having the Unit 2 DC Bus cross-tied to Unit 1. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 52919 | 21 August 2017 02:08:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Assessment Capability Due to Non-Functional Tsc Ventilation | On August 20, 2017 at 2108 hours CDT, Byron Station Technical Support Center (TSC) ventilation system supply fan (0VV23C) was identified as non-functional. This failure affects the ability of the TSC ventilation system to maintain adequate radiological habitability in the event of an emergency with an airborne radiological release. All other capabilities of the TSC are unaffected by this emergent condition. Currently troubleshooting/investigation is being performed. This condition is considered a major loss of emergency assessment capability and is reportable under 10CFR50.72(b)(3)(xiii). If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures. If the TSC becomes uninhabitable, the Station Emergency Director will relocate the TSC staff to an alternate TSC location in accordance with applicable procedures. This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the potential loss of an emergency response facility because of the unavailability of the ventilation system. An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified.
TSC ventilation was returned to service at 1150 CDT on 8/23/17. The licensee has notified the NRC Resident Inspector. Notified R3DO (Dickson). | ||
| ENS 52592 | 5 March 2017 14:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Planned Ultrasonic Test on Previously Repaired Reactor Vessel Penetration Did Not Meet Acceptance Criteria | On 3/5/2017 at 0800 (CST), during the Byron Station Unit 1 refueling outage, it was determined that the results of a planned Ultrasonic (UT) examination performed on a previous repaired penetration of the reactor vessel head did not meet applicable acceptance criteria. These indications are not in the reactor coolant pressure boundary; however they are very near the previously repaired J-groove weld. The indication will be addressed prior to returning the vessel head to service. The examination was being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1 to ensure the structural integrity of the reactor vessel head pressure boundary. This event is being reported under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The licensee has notified the NRC Resident Inspector. Exelon Generation Company, LLC is also notifying the NRC Division of Component Integrity or its successor, by use of this ENS Report, of changes in indication(s) or findings of new indications(s) in the penetration nozzle or J-groove weld beneath a seal weld repair, or new linear indications in the seal weld repair, prior to commencing repair activities. See also EN #52591
Byron Event Notification 52952, made on March 05, 2017, reported two indications on one reactor head penetration. On March 06, 2017, ultrasonic examination identified two additional indications on the same penetration nozzle, (P-76). Additionally, two indications were identified on a second penetration nozzle (P-74), and one indication was identified on a third penetration nozzle (P-77). The indications on these three penetrations are all of the same type and none are within the reactor vessel head pressure boundary. This event is being reported under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The NRC Resident Inspector has been notified. Exelon Generation Company, LLC is also notifying the NRC Division of Component Integrity or its successor, by use of this ENS Report, of changes in indication(s) or findings of new indications(s) in the penetration nozzle or J-groove weld beneath a seal weld repair, or new linear indications in the seal weld repair, prior to commencing repair activities. Notified R3DO (Jeffers) | 05000454/LER-2017-001 | |
| ENS 52591 | 3 March 2017 23:05:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Previous Overlay Repair Did Not Meet Acceptance Criteria | On 3/3/2017 at 1705 (CST), during the Byron Station Unit 1 refueling outage, it was determined that the results of a planned Liquid Penetrant (PT) examination performed on a previous overlay repair of the reactor vessel head did not meet applicable acceptance criteria. The penetration overlay weld indication will be addressed prior to returning the vessel head to service. The examination was being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1 to ensure the structural integrity of the reactor vessel head pressure boundary. Ultrasonic examinations have been performed and no ultrasonic indications were identified. This event is being reported under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The licensee has notified the NRC Resident Inspector. | 05000454/LER-2017-001 | |
| ENS 52295 | 12 October 2016 18:38:00 | 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 Spurious Feedwater Isolation | At 1338 CDT, a spurious electrical fault on a unit substation motor control center caused a spurious feedwater isolation and required a manual reactor trip on Unit 2 Byron Station due to lowering steam generator level. Auxiliary feedwater automatically actuated to supply feedwater flow and is maintaining steam generator level within procedurally required levels. The motor control center fault resulted in feedwater isolations on two of four steam generators. All rods inserted during the trip. Decay heat is being removed via the steam dumps to condenser. Plant response to the trip was uncomplicated. Unit 2 is in a normal shutdown electrical lineup with power available from offsite. The grid is stable. Byron Station Unit 1 is at 95% power and stable and was unaffected by the transient on Unit 2. The cause of the motor control center fault is under investigation. The licensee has notified the NRC Resident Inspector and will be issuing a press release concerning the manual reactor trip. | Steam Generator Feedwater Auxiliary Feedwater | 05000455/LER-2016-001 |
| ENS 52148 | 2 August 2016 14:34:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Watertight Door Discovered Open and Unattended | On August 2, 2016 at 0934 CDT, Byron Operators entered and exited 1BOL 8.1 Conditions B and F for the 1A and 1B Diesel Generators due to 0DSSD192 (1B DOST Watertight Door) being discovered open and unattended. 0DSSD192 was closed within 5 minutes of discovery. 0DSSD192 protects the DOST (diesel oil storage tank) transfer pumps from the effects of a postulated failure of a Circulating Water expansion joint at the condenser waterboxes in the Turbine Building. An open watertight door associated with one DOST has the potential of making both Diesel Generators inoperable. This event is reportable per 10 CFR 50.72(b)(3)(v)(D) for any event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector.
The purpose of this report is to retract a previous report made on August 2, 2016 at 1755 EDT (EN 52148). Notification of the event to the NRC was initially made for a condition where the station determined that an open watertight door associated with one DOST (Diesel Oil Storage Tank) had the potential to make both Diesel Generators inoperable, and the condition was reported under 10 CFR 50.72(b)(3)(v)(D) for any event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to mitigate the consequences of an accident. Upon further investigation, Byron Station has concluded that the 1A Diesel Generator was never inoperable, and therefore, no loss of safety function occurred. Based on this, the prior ENS notification is being retracted. The NRC Senior Resident Inspector has been notified. Notified R3DO (Dickson). | ||
| ENS 52100 | 18 July 2016 20:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Offsite Communications Capability | During testing of the Everbridge Emergency Response Organization (ERO) notification system, it was found that the system may not have notified corporate ERO members within the required ten minutes from system initiation. This constitutes a loss of offsite communication capabilities. The Everbridge vendor has since restored and successfully tested the system. Additionally, the site successfully conducted an ERO call-in drill using the same Everbridge notification system. The Emergency Response Data System was not affected. This event is reportable per 10 CFR 50.72(b)(3)(xiii) as a loss of communications capability. The NRC Resident Inspector has been notified. The requirement is to have all ERO personnel receive the page within ten minutes. Compensatory measures were instituted while the system was not functional. The system was returned to service at 2100 CDT following repair and testing. | Emergency Response Data System | |
| ENS 51958 | 25 May 2016 18:00:00 | 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 | Discovery of Non-Conforming Conditions During Tornado Hazards Analysis | On May 25, 2016, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Byron Station identified non-conforming conditions in the plant design such that specific TS equipment on both units is considered to not be adequately protected from tornado missiles. Each of the following reportable conditions is postulated by tornado missiles entering openings through the wall that separates the Auxiliary Building and the Turbine Building: The following items could be impacted by tornado missiles entering the 0A and 0B (common) MCR (main control room) turbine building intake openings: Main control room pressurization, main control room ductwork and dampers, chilled water to the VC (control room ventilation) coils, main control room radiation monitors. The following items could be impacted by tornado missiles entering the Division 11 and 21 MEER (miscellaneous electrical equipment rooms) rooms: Exhaust from the MEER (affects room cooling and MCR pressure), exhaust from a battery room, conduits and cabinets associated with the battery chargers and DC bus. The following items could be impacted by tornado missiles entering the Division 12 and 22 MEER rooms: Exhaust from the MEER (affects room cooling and MCR pressure), MEER supply fan and ductwork, battery room exhaust, and the instrument inverter cabinets. The RWST (refueling water storage tank) roof access opening Bilco hatch is fabricated from sheet metal that is not designed to prevent all postulated tornado missiles from entering the tank. The tank pressure boundary is 24" thick concrete and is designed to withstand an external tornado missile impact. Thus a missile that enters the tank will not adversely impact the tank pressure boundary. The following items could be impacted by tornado missiles entering the RWST roof access: The 6" RWST recirculation pipe, 3" overflow pipe, and 24" suction pipe. This piping is located inside the tank and they are approximately 130 degrees around the tank away from the hatch opening. This condition creates a potential LOSF (loss of safety function) with the Byron Essential Service Water Cooling Towers (UHS) (ultimate heat sink) with the discovery that the power and control cables to four of eight cooling tower fans can be damaged by tornado missiles penetrating through wall openings. 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. These conditions are being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The licensee has notified the NRC Resident Inspector. | Service water | |
| ENS 51931 | 15 May 2016 16:18:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Excessive Unidentified Leakage During Plant Startup | Unit 2 experienced RCS Leakage, potentially pressure boundary leakage, or Unidentified Leakage greater than 10 gpm for (greater than) 15 minutes. Reactor Vessel Flange Temperature High Leakoff alarm was received. This met threshold for declaration of an Unusual Event at 1118 CDT per MU6. Unit 2 is currently in MODE 3. Investigation is in progress to identify specific leakage location. The plant is stable, leakage indicates about 32 gpm and the startup has been stopped. There is no impact on Unit 1. The licensee notified the NRC Resident Inspector and State authorities. Notified DHS SWO, FEMA Ops Center, DHS NICC,. Notified FEMA National Watch and Nuclear SSA via email.
At 1459 CDT, the conditions under MU6 are no longer met. The site has terminated the Unusual Event. In addition, a press release will be made of this event. Unit 2 is stable and in Mode 3 pending further evaluation. The leak stopped when a loop drain isolation valve was closed. The licensee notified the NRC Resident Inspector and State and local authorities. Notified the R3DO (Duncan), NRR EO (Morris), IRD MOC (Grant), DHS SWO, FEMA Ops Center, and DHS NICC. Notified the FEMA National Watch and Nuclear SSA via email. | ||
| ENS 51806 | 18 March 2016 14:37:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Confirmed Positive Test for Alcohol | In accordance with 10 CFR 26.719(b)(2)(ii), this notification reports a non-licensed contracted employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The employee's access to the plant has been suspended. The licensee has notified the NRC Resident Inspector. | ||
| ENS 51772 | 7 March 2016 10:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Involving Diesel Driven Auxiliary Feedwater Pump Air Intakes | The Auxiliary Feedwater (AF) system at Byron automatically supplies feedwater to the Steam Generators (SG) to remove decay heat from the Reactor Coolant System following a loss of normal feedwater supply. The AF System consists of a motor driven pump (A) and a diesel driven pump (B) configured into two trains for each unit. Each pump provides 100% of the required AF capacity to the SGs as assumed in the accident analysis. One pump at full flow is sufficient to remove decay heat and cool the unit to Residual Heat Removal (RHR) entry conditions. The diesel driven AF pump is powered from an independent diesel whose combustion air intake is located in the Seismic Category II (non-seismically qualified) Turbine Building but the diesel and pump are located in the Seismic Category I (seismically qualified) Auxiliary Building. During the ongoing NRC Component Design Basis Inspection at the sister Braidwood Station, inspectors asked about the acceptability of the diesel combustion air intake being located in the non-seismic Turbine Building. During the review of available documentation related to the AF diesel engine combustion air intake, it was identified that the documentation did not support operation of the diesel with High Energy Line Break (HELB) environmental conditions in the Turbine Building. This has been reviewed and determined to be applicable to Byron Station Units 1 and 2. Specifically, prior evaluations did not account for air displacement by steam release during the event. After running different models for the Turbine Building HELB, diesel driven AF pump operability was supported for all but the Main Feedwater (FW) HELB. For the FW HELB, the best air density obtained failed to remain above the required levels deemed acceptable for engine operation and remained suppressed for extended periods of time. Additional efforts to qualify the FW piping in the Turbine Building for an Operating Basis Earthquake (OBE) to eliminate this piping from HELB considerations were not successful. This condition applies to both Units 1 and 2 but does not affect the motor driven AF pumps. This event does not constitute a loss of safety function at the point of discovery because the Byron opposite train motor driven AF pumps were operable on both Units 1 and 2. This event 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.' The licensee has notified the NRC Resident Inspector. The licensee entered a 72-hour Action Statement and engineering is analyzing the issue. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Residual Heat Removal | 05000454/LER-2016-001 |
| ENS 51633 | 3 January 2016 05:40:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Inadvertant Siren Actuation | On 01/02/2016 at 2340 (CST), personnel at the Ogle County 9-1-1 call center inadvertently actuated the emergency response sirens associated with Byron Generating Station multiple times. The emergency response sirens were secured at approximately 2350. This event is being reported under 10CFR50.72(b)(2)(xi) for, 'any event or situation, related to the health and safety of the public or on site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The licensee has notified the State of Illinois and the NRC Resident Inspector. All sirens are operable and available if needed. | ||
| ENS 51436 | 1 October 2015 14:06:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Mode Change with Turbine Trip Leads Lifted | At 0906 (CDT), it was determined that U1 (Unit 1) was in a condition that could have prevented fulfillment of the turbine trip safety function and TS (Tech Spec) 3.0.3 was entered. Leads had been lifted to disable the turbine trip function on both SSPS (Solid State Protection System) trains while U1 was in Mode 4 (which is outside the mode of applicability). However, at 0059 (CDT), U1 entered Mode 3 with these leads still lifted. In Mode 3, both trains of the turbine trip function are required to be operable per TS 3.3.2. The turbine was subsequently tripped at 0932 (EDT) and the leads were re-landed enabling the turbine trip function at 0946 (CDT), TS 3.0.3 was subsequently exited. This condition is being reported in accordance with 10 CFR 50.72 (b)(3)(v)(D) for an event or condition that could have prevented the fulfillment of a safety function to mitigate the consequences of an accident. The Byron NRC site Resident Inspector has been notified of this condition. The licensee has notified the State of Illinois. | 05000454/LER-2015-006 | |
| ENS 51410 | 19 September 2015 01:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Vessel Head Penetration Repair Did Not Meet Acceptance Criteria | On 9/18/2015 at 20:00 (CDT), during the Byron Station Unit 1 refueling outage, it was determined that the results of a planned Liquid Penetrant (PT) examination performed on a previous overlay repair of the reactor vessel head did not meet applicable acceptance criteria. The penetration requires repairs prior to returning the vessel head to service. These indications are not in the reactor coolant pressure boundary; however they are very near the previously repaired J-groove weld. The examination was being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to ensure the structural integrity of the reactor vessel head pressure boundary. No ultrasonic indications have been identified at this time. Repairs are currently being planned in accordance with the ASME Code of Record. The repairs will be completed prior to returning the vessel head to service. This event is being reported under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The licensee has notified the NRC Resident Inspector. Per regulatory commitment, Exelon Generation Company, LLC (EGC) is notifying NRC staff of the Division of Component Integrity or its successor, of changes in indication(s) or findings of new indication(s) in the penetration nozzle or J-groove weld beneath a seal weld repair, or new linear indications in the seal weld repair, prior to commencing repair activities. The original indications that led to the overlay repairs were discovered during ultrasonic testing and were reported to the NRC and assigned EN46686 and EN48311. | 05000454/LER-2015-005 | |
| ENS 51412 | 17 September 2015 05:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Control Room Area Radiation Monitors Not Able to Provide Radiation Level Information | On 9/17/2015 at 0030 (CDT), during the Byron Station Unit 1 refueling outage, Main Control Room area radiation monitors were removed from service to support an electrical bus outage. During this time, the Main Control Room area radiation monitors were not able to generate Main Control Room annunciation or provide area radiation level information necessary for Emergency Action Level (EAL) threshold determination until the area radiation monitor is restored which is scheduled for 9/20/15. This event is being reported under 10 CFR 50.72(b)(3)(xiii) for 'Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability.' The licensee has notified the NRC Resident Inspector. The licensee discovered this issue at 2145 CDT on 9/19/15. The licensee established compensatory measures upon discovery of the issue. | ||
| ENS 51401 | 17 September 2015 03:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Seismic Monitoring Instrumentation During Planned Maintenance | At 2200 CDT on September 16, 2015, Byron Station's Seismic Instrumentation will be removed from service to support an electrical bus outage. During this time, the seismic instrumentation will not be able to generate Main Control Room annunciation or provide ground acceleration information necessary for Emergency Action Level (EAL) threshold determination until the seismic instrumentation is restored, which is scheduled for 1700 CDT on September 19, 2015. Since the duration of maintenance activity may last greater than 72 hours, with viable compensatory measures in place and communicated to applicable Emergency Response Decision Makers, this condition will result in a Loss of Emergency Assessment Capability while the Seismic Instrumentation is out of service and results in a reportable condition in accordance with 10 CFR 50.72(b)(xiii). The Licensee has notified the NRC Resident Inspector and informed the State of Illinois Resident Engineer . | ||
| ENS 51335 | 20 August 2015 22:55:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Condition That Could Prevent Pressurizer Porv Block Valves from Operating | On 8/20/2015 at 1755 (CDT), a design flaw was discovered with the pressurizer power operated relief valve (PZR PORV) block valve control circuitry. Specifically, the circuit deficiency for which a design basis fire in the Main Control Room (MCR) or cable spreading room could prevent the PZR PORV block valves from being closed from the local control switch at their associated motor control center (MCC). Engineering has reviewed this issue and determined that a potential fire induced ground in the MCR or cable spreading room could clear the associated control power fuses which would prevent the block valves from operating at the local control switch. These valves are considered to form a High/Low pressure interface which requires postulating a proper polarity DC cable to cable fault. Engineering has reviewed the circuit design and cable routing associated with PORVs 1(2)RY455A and 1(2)RY456 and determined that their associated cables are routed with other DC circuit cables in the MCR control board and cable spreading room raceways, such that this postulated fault could potentially cause spurious opening of one of the PORVs even after the control power fuses have been removed as directed by the station abnormal operating procedures for control room inaccessibility. This identified block valve circuit deficiency prevents the credited safe shutdown action of locally closing the block valves to mitigate the spurious operation of a PORV. Hourly fire watches of the affected MCR and cable spreading room fire zones have been implemented. In addition, the MCR is continuously staffed and the affected cable spreading room fire zones are equipped with detection and automatic suppression. This event is being reported under 10CFR50.72(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. The licensee has notified the NRC Resident Inspector.
During the extent of condition review, an additional design deficiency was identified with respect to the PZR PORV and PZR PORV Block valves control circuitry. Specifically, the current mitigating strategy for removing PZR PORV control power fuses does not adequately prevent a PZR PORV from spuriously opening due to fire induced hot short. Furthermore, local actions to close the associated PZR PORV block valve at the motor control center (MCC) may not be effective because the MCC may not have electrical power during the design basis fire. Therefore, the credited safe shutdown action to remove the PZR PORV control power fuses does not prevent the PZR PORV from spuriously opening during design basis fires in some of the upper and lower cable spreading room fire zones. The affected Fire Zones are the same upper and lower spreading rooms previously identified and fire watches of the affected areas remain in place. The NRC Resident Inspector has been notified. Notified the R3DO (Skokowski). | 05000454/LER-2015-004 | |
| ENS 51119 | 4 June 2015 13:39:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness-For-Duty Report Involving a Non-Licensed Supervisory Employee | A non-licensed supervisory employee had a confirmed positive for alcohol during a for-cause fitness-for-duty test. The employee's access to the plant has been suspended. The licensee notified the NRC Resident Inspector. | ||
| ENS 50988 | 16 April 2015 10:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Scheduled Maintenance Affecting Technical Support Center Hvac | On April 16, 2015, Byron Station will remove part of the Technical Support Center (TSC) emergency ventilation system from service to facilitate necessary surveillance work on the fire protection system. This work is expected to last approximately 4 hours. This maintenance affects the ability of the TSC ventilation to maintain adequate habitability during the duration of an emergency. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures. If the TSC becomes uninhabitable, the Station Emergency Director will relocate the TSC staff to an alternate TSC location in accordance with applicable procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the potential loss of an emergency response facility because of the unavailability of the ventilation system. An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified.
Maintenance has been completed and the TSC emergency ventilation system was restored to service at 1300 CDT on 4/16/15. The NRC Resident Inspector will be notified. Notified the R3DO (McCraw). | ||
| ENS 50881 | 12 March 2015 14:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of the Seismic Monitoring Computer | In accordance with 10 CFR 50.72(b)(3)(xiii), this notification reports a loss of Emergency Preparedness assessment capability with the unplanned inoperable condition for the Byron seismic monitor. Specifically, the seismic monitor was declared non-functional at 0345 CST on March 7, 2015 following an unplanned loss of the seismic monitoring central computer. This condition adversely impacted the capability to perform an ALERT EAL (HA4) assessment in accordance with the Radiological Emergency Plan Annex. The loss of assessment capability is reportable to the NRC within 8 hours of the discovery in accordance with 10 CFR 50.72(b)(3)(xiii) The seismic recorder was reset by Plant Engineering at 0440 CST on March 8 2015, which restored the seismic monitoring system to full capability. The unplanned non-functional condition of the seismic monitor was entered into the Byron CAP (Corrective Action) Program when the condition was discovered. Initially it was determined that the condition did not meet the requirements for an ENS notification as it was concluded that it did not substantially impair Byron station's emergency assessment capability in the event of an earthquake. Upon further review, at 0900 CDT on March 12, 2015, it was concluded that an ENS notification was warranted because the monitor was specifically Cited in the Emergency Action Level (EAL) HA4 threshold for identifying an ALERT due to an OBE (Reference procedure EP-AA-1002, Addendum 3, 'Emergency Action Levels for Byron Station'). A follow-up written notification is not required for this notification under 10 CFR 50.73. The NRC Resident Inspector has been notified." The State of Illinois Resident Inspector has been notified.
Upon further review of this event for the time period of 3/07/12 through 3/17/15, Byron Station has identified six previous occurrences where the Seismic Monitor was declared non-functional, which impacted the capability to perform an ALERT EAL (HA4) assessment in accordance with the Radiological Emergency Plan Annex. These occurred on November 6, 2014; April17, 2013; January 2, 2013; October 10, 2012; July 18, 2012 and July 9, 2012. The NRC Resident Inspector has been notified. Notified the R3DO (Roach). | ||
| ENS 50859 | 3 March 2015 17:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Main Power Transformer Bushing Short | In accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A), this notification reports an automatic reactor trip on Byron Unit 1. The trip occurred following a trip of the Byron Unit 1 East Main Power Transformer (MPT). Initial indications appear that the MPT trip was caused by a large (~ 5 foot) section of ice that fell from a bus bar over the 1E MPT. This ice shorted out a MPT bushing, resulting in the unit trip. Reactor operators performed a manual start of the Auxiliary Feedwater System in response to the unit trip. All other safety systems responded as expected. The plant trip occurred at 1101 CST on March 03, 2015. Unit 1 is presently in Mode 3 and stable. Unit 2, the opposite unit, is operating at 100% power and stable. This condition was entered into the Byron CAP Program. An investigation is in progress to determine the extent of required repairs, if any, required prior to unit restart. This event resulted in the actuation of the Reactor Protection System with a subsequent Reactor Trip and therefore, requires notification to the NRC within 4 hours of discovery in accordance with 10 CFR 50.72(b)(2)(iv)(B). This event resulted in the manual actuation of the Auxiliary Feedwater System and therefore, requires notification to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(iv)(A). This ENS call will be followed up with a Licensee Event Report (LER) within 60 days. All rods inserted during reactor trip, offsite power and emergency power sources are currently available and decay heat is being removed via the startup feedwater systems. No safety relief valves lifted as a result of the transient. The NRC Resident Inspector and the State of Illinois were notified. | Feedwater Reactor Protection System Auxiliary Feedwater Safety Relief Valve | |
| ENS 50517 | 8 October 2014 00:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Ultrasonic Testing Reveal Degraded Condition on One Control Rod Drive Mechanism Penetration | In accordance with 10 CFR 50.72(b)(3)(ii)(A), this notification reports a degraded condition on Byron Station Unit 2 that occurred on October 7, 2014, when it was determined that the results of planned ultrasonic testing (UT) examinations performed on one CRDM penetration of the reactor vessel head would not meet the applicable acceptance criteria. Byron Station Unit 2 is presently in day eight of a refueling outage. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1 to ensure the structural integrity of the reactor vessel head pressure boundary. The UT examinations continue for the remaining CRDM head penetrations. The repairs are currently being planned and will be completed prior to unit start-up. The NRC Resident Inspector has been notified of this condition. | Control Rod | |
| ENS 49919 | 15 March 2014 16:02:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Momentary Loss of Shutdown Cooling Due to a Loss of Offsite Power | (Unit) 1 experienced a loss of offsite power (LOOP). (Unit) 1 station aux transformers (SATs) tripped. 1A and 1B emergency diesel generators (EDGs) started and properly restored power to ESF (engineered safety feature) bus 141 and 142. Safe shutdown loads (1A/B SX (essential service water) pumps and 1A/B CC (component cooling) pumps) properly started when signaled by the EDG sequencers. The 1A residual heat removal pump was manually restarted at 1105 (CDT) to restore shutdown cooling. There was no impact to RCS temperature. Additionally, a containment isolation signal was actuated when containment radiation monitors which are powered from the ESF buses briefly lost power until restored by the EDGs. All related equipment realigned as expected except for the 0A VC (control room) chiller that didn't start when required on the EDG sequencer. The cause of the LOOP and the 0A VC chiller issue is under investigation. Fuel moves were in progress and stopped safely. All fuel bundles are in a safe location. Unit 2 was not affected and remains in mode 1. The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator Shutdown Cooling Residual Heat Removal | |
| ENS 49599 | 3 December 2013 04:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center (Tsc) Non-Functional Due to Planned Maintenance | On December 2, 2013, activities are scheduled that will render the Technical Support Center (TSC) non-functional by removing the normal and emergency ventilation system from service. These activities are being performed in support of planned preventive maintenance. In preparation for these normal and emergency ventilation system outages, the TSC emergency responders were notified that if an emergency occurred during this outage the Emergency Coordinator and the TSC staff involved with classification, notification and PARS should report to the Work Execution Center. All other TSC personnel should report to the Operational Support Center. The duration of this TSC outage is expected to be less than 36 hours. The NRC Operations Center will be provided an update to this notification when the TSC normal and emergency ventilation is restored. This 8 hour notification in accordance with 10 CFR 50.72(b)(3)(xiii). The licensee notified the NRC Resident Inspector.
The TSC is fully functional. The licensee will notify the NRC Resident Inspector. Notified R3DO (Riemer). | ||
| ENS 49375 | 21 September 2013 20:31:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Commercial and Ens Communications Due to an Area Phone Outage | At 1531 CDT, Byron Station was contacted via NARS (Nuclear Accident Reporting System) phone line by the Illinois Emergency Management Agency (IEMA) that an issue exists for the phone systems in the Byron area. Byron subsequently contacted lEMA via the NARS phone and was able to communicate with lEMA. At 1536 CDT, the ENS phone was verified to not be functioning and at 1541 CDT NRC Headquarters was contacted via Byron Main Control Room cell phone to report the failure of the ENS and commercial phones. The Byron Station Main Control Room cell phone is functioning normally. The loss of ENS and commercial phone capability is reportable under 10CFR50.72(b)(3)(xiii). The cause of the loss of phone capability is not known but is under investigation and an update will be provided upon restoration. ERDS has been verified as operable and the licensee is transmitting data. The licensee has notified the NRC Resident Inspector. | ||
| ENS 49138 | 20 June 2013 16:18:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Ventilation Removed from Service for Maintenance | On June 20, 2013 at 1118 CDT, Byron Station removed part of the Technical Support Center (TSC) ventilation (i.e., 0VV23C) system from service to facilitate necessary maintenance on the supply fan. This work is expected to last approximately 2 hours. This maintenance affects the ability of the TSC ventilation to maintain adequate habitability during the duration of an emergency. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures. If the TSC becomes uninhabitable, the Station Emergency Director will relocate the TSC staff to an alternate TSC location in accordance with applicable procedures. This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the potential loss of an emergency response facility because of the unavailability of the ventilation system. An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified (by the Licensee).
Maintenance has been completed and the TSC ventilation system has be restored to service. The licensee will notify the NRC Resident Inspector. Notified R3DO (Daley). | ||
| ENS 49137 | 20 June 2013 11:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Sulfuric Acid Spill Requiring Offsite Notification | In the owner controlled area, the plant has a tank of sulfuric acid (93% concentration, UN # 1830) for chemical control of the Circulating Water system used in the non-safety related electrical power production portion of the plant. At 0600 (CDT), a routine inspection of the tank found acid within the berm surrounding the tank. A leak was discovered in the berm and sulfuric acid leaked onto the surrounding area and nearby drainage ditch. The quantity of leaked acid has been estimated to be 120 gallons. The source of the leak has been determined to be a pipe on the tank and it has been isolated with a closed valve. The tank level has been stable for 4 hours. The leaked acid is contained within approximately 70 linear feet of the (limestone) ditch and has not left the site. On-site operations and chemistry personnel are neutralizing the acid with soda ash and a tanker truck was contracted to arrive on site this morning to empty the acid tank contents until a repair can be accomplished. The leak was discovered at 0600 and isolated about 0630 CDT. The licensee notified the NRC Resident Inspector. | Circulating Water System | |
| ENS 48838 | 21 March 2013 00:51:00 | 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 Loss of Generator Stator Cooling Water | At 1951 CDT on March 20, 2013, Byron Unit 2 Reactor was manually tripped due to the loss of all Generator Stator Cooling Water. 2BEP-0, 'Reactor Trip or Safety Injection Unit 2' was entered and a transition was made to 2BEPES 0.1, 'Reactor Trip Response Unit 2.' The auxiliary feedwater pumps automatically actuated upon the expected low steam generator level. Upon the trip, it was noted that a Digital Rod Position Indication System Urgent Failure occurred with a General Warning on Control Rod position M12. Indication for the Train 'B' Reactor Trip breaker was lost. All Control Rods inserted upon Reactor trip and the Train 'B' Reactor trip breaker was locally verified open. The plant is in its normal shutdown electrical lineup. No safeties or reliefs lifted during the event. There was no impact on unit-1. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater Stator Water Control Rod | |
| ENS 48572 | 7 December 2012 20:35:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Employee Had a Confirmed Positive for Alcohol | A non-licensed, non-supervisory employee had a confirmed positive for alcohol during a for-cause fitness-for-duty test. The employee's access to the plant has been suspended. The licensee notified the NRC Resident Inspector. | ||
| ENS 48511 | 1 October 2012 20:03:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Telephonic Notification of Invalid System Actuation | In accordance with 10CFR50.73(a)(2(iv)(A), this telephone notification reports an invalid actuation of the Unit 1 train A (1A) Emergency Diesel Generator (DG) on October 1, 2012, at 1403 hours. At the time of the event Unit 1 was in Mode 5, Cold Shutdown. The 1A DG was undergoing a surveillance procedure that verifies a safety injection override signal when a technician actuated the incorrect under voltage relay which resulted in the inadvertent start of the 1A DG. The engine functioned successfully and ran unloaded, as expected. The cause is attributed to the technicians failing to verify they were actuating the correct relay. This condition was entered into the corrective action program. The licensee notified the NRC Resident Inspector and the Illinois Department of Nuclear Safety. | Emergency Diesel Generator | |
| ENS 48311 | 15 September 2012 00:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Indications on Reactor Vessel Head During Dye Penetrant Test | On September 14, 2012, during the Byron Station Unit 1 refueling outage, it was determined that the results of planned Liquid Penetrant (PT) examinations performed on two previous overlay repairs of the reactor vessel head do not meet applicable acceptance criteria. Both penetrations require repairs prior to returning the vessel head to service. These indications are not in the reactor coolant pressure boundary; however they are very near the previously repaired J-groove weld. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to ensure the structural integrity of the reactor vessel head pressure boundary. No ultrasonic indications have been identified at this time. Repairs are currently being planned, which will include buff of the rejectable area and retest, and will be completed prior to returning the vessel head to service. If retest of the rejectable areas is unacceptable, then additional repairs will be required prior to returning the vessel head to service. This condition is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indications did not meet applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. Any further examination failures and repairs will be updated under this ENS Notification. The NRC Resident Inspector has been notified. The original indications that led to the two overlay repairs were discovered during ultrasonic testing and were reported to the NRC and assigned EN #46686. | 05000454/LER-2012-004 | |
| ENS 48290 | 10 September 2012 14:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Plant Process Computer Removed from Service for Maintenance | At 0900 CDT on September 10, 2012, the Unit 1 Plant Process Computer (PPC) was removed from service for a planned replacement in the current Unit 1 Refueling Outage. The Unit 1 PPC feeds the Safety Parameter Display System (SPDS) used in the Main Control Room (MCR) and the Technical Support Center (TSC). The Unit 1 PPC also feeds the Emergency Response Data System (ERDS). The Unit 1 and Unit 2 PPCs also feed the Plant Parameters Display System (PPDS) used in the MCR, TSC and Emergency Operations Facility (EOF). Meteorological data will remain available. The dose assessment program will remain functional as the Unit 2 PPC will be capable of providing the necessary data through PPDS to run the program. The dose assessment program is not affected by the Unit 1 PPC being out of service. As compensatory measures, a proceduralized backup method to fax or communicate via a phone circuit applicable data to the NRC, TSC, and EOF exists. There is no impact on the Emergency Notification System (ENS) or Health Physics Network (HPN) communication systems. The new Unit 1 PPC is scheduled to be functional on September 17, 2012. However, based on the Mode Unit 1 will be in, this will limit the number of points that would provide usable data. The Unit 1 PPC will be tested as Mode changes occur. The Unit 1 PPC is planned to be declared functional by Mode 2. A follow-up ENS call will be made once the Unit 1 PPC is declared functional. The loss of SPDS and ERDS is a 'major loss of assessment capability' and is reportable under 10CFR50.72(b)(3) (xiii). The NRC Senior Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS call.
As of 1245 CDT on October 3,2012, the Unit 1 PPC is considered operational with respect to the Safety Parameter Display System (SPDS), Plant Parameter Display System (PPDS), and Emergency Response Data System (ERDS). Therefore, a major loss of assessment capability no longer exists on Unit 1. The Byron EP manager contacted the NRC ERDS Center on 10/02/2012 to conduct an ERDS test for Unit 1 to ensure the data was being satisfactorily sent to the NRC. Unit 1 and Unit 2 ERDS data to NRC was tested satisfactory lAW EP-AA-124-F�01. The NRC Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS update. Notified the R3DO (Dave Passehl) | ||
| ENS 48278 | 5 September 2012 17:25:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification of Sodium Hypochlorite Spill | At 1225 on 9/5/12, the Main Control Room received a report of a Sodium Hypochlorite spill estimated at 100 gallons at the Byron Circulating Water Pump House. The spill was caused by a leak from the hose of the Chemical Delivery truck. A spill of greater than 80 gallons of Sodium Hypochlorite is considered a reportable quantity. The spill was reported to the Illinois Emergency Management Agency (IEMA) in accordance with 29 IAC 430.30 and 35 IAC 750.304, and the National Response Center in accordance with 40 CFR 302.6. Notification to IEMA was performed at 1235 on 9/5/12, and notification to the National Response Center was performed at 1245 on 9/5/12. The notifications to other government agencies is reportable to the NRC in accordance with 10 CFR 50.72(b)(2)(xi). The licensee has notified the NRC Resident Inspector. | ||
| ENS 47708 | 28 February 2012 23:31:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Due to Loss of Offsite Power | Byron Unit 1 declared an Unusual Event due to a loss of offsite power for greater than 15 minutes (EAL MU-1). The 1A and 1B emergency diesel generators auto-started and automatically restored power to ESF (Engineered Safety Feature) Busses 141 and 142. The 1A Auxiliary Feedwater pump auto-started due to an undervoltage signal to ESF bus 141. The 6.9 KV Busses (Non-ESF) and 4.16 KV Busses (Non-ESF) auto transferred to the Unit Auxiliary Transformers and Unit 1 remained online. The ESF reserve feed breakers are available and in the process of transferring ESF loads over to Unit 2 via the ESF cross-ties. This will allow shutdown of the 1A and 1B diesel-generators and allow alignment to standby status. Switchyard repairs are being initiated. The licensee notified the State of Illinois, Ogle County, Oregon and Byron municipalities and the NRC Resident Inspector. The licensee plans on issuing a news release. Unit 2 remained online and was unaffected throughout the event. Notified Nuclear SSA and NICC vie email.
Termination of Unusual Event. Repairs to switchyard components are complete. Offsite power has been restored to Unit One System Auxiliary Transformers. Unit One 4 KV ESF busses have been re-energized from Unit One System Auxiliary Transformers and are no longer cross tied to Unit Two System Auxiliary Transformers. This restores the normal electrical lineup to the ESF busses. The entry conditions for MU-1 are no longer met. The licensee has notified the NRC Resident Inspector, the State of Illinois, Ogle County, Oregon and Byron municipalities. Notified R3DO (Stone), NRR EO (Lee), IRD (Morris), and other FEDS (DHS SWO, DHS NICC, FEMA, and NuclearSSA via email). | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 47644 | 6 February 2012 23:19:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Rps Actuation Due to Turbine Trip | Unit 2 had been connected to the grid at 1431 CST today following a recent forced outage and was in the process of power ascension in accordance with plant procedures. At approximately 25% reactor power (approximately 240 MWe), the operators were transferring main feed flow to the steam generator lower nozzle through the feedwater isolation valves, (i.e. 2FW009A, B, C, D). 2FW009C was the first valve to be opened due to previously experienced problems with this valve being stuck in the seat. No issues were experienced during the opening of 2FW009C. Upon opening of 2FW009C, the 2C steam generator level began to rise as expected. The operators throttled back feedwater flow to control steam generator level. However, the 2C steam generator level increased to the High-High level setpoint of 80.8% (p-14). Since the reactor was below 30% (P-8), no automatic reactor trip signal was generated. However, the turbine automatically tripped, a feedwater isolation signal was initiated, and 2C Main Feedwater pump trip occurred as designed. With no main feedwater flow available, the operators manually tripped the reactor and entry into procedure 2BEP 0, Reactor Trip or Safety Injection Unit 2 was entered. The operators then manually started the 2A and 2B Auxiliary Feedwater pumps to supply water to the steam generators prior to reaching the Low-Low steam generator level setpoint of 36.3%. Transition from 2BEP 0 to 2BEP ES-0.1, Reactor Trip Response, was completed and the emergency procedures were exited. Unit 2 is being maintained in a stable condition in Mode 3. The NRC Resident Inspector has been informed. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 47636 | 3 February 2012 06:00:00 | Other Unspec Reqmnt | VOLUNTARY REPORT - DESIGN VULNERABILITY IN 4.16kV BUS UNDER-VOLTAGE SCHEME On January 30, 2012, a design vulnerability was discovered at Byron and Braidwood stations in the Engineered Safety Feature 4.16kV bus under-voltage protection scheme for Byron Station Units 1 and 2. Specifically a voltage unbalance created by an open circuit of either the A or C phase from the offsite grid to the System Auxiliary Transformers (SAT) is not designed to actuate the protective relays on the 4.16kV safety bus that provides isolation from the offsite grid and the automatic start and loading of the emergency onsite diesel generators. Two under-voltage relays are provided on each 4.16kV safety bus, which are combined in a two out of two logic to generate a loss of power signal. The relays are sensing voltage between two phases (i.e., A&B and B&C). An open circuit condition on the C phase or the A phase would not satisfy the two out of two logic. This condition results in both 4.16kV safety buses remaining energized with a bus undervoltage situation and results in equipment protective devices actuating from over-current conditions. This configuration is a non-conforming condition in that the design of the under-voltage relays and logic was intended to identify degraded grid conditions, not loss of a single phase. With an open circuit on the A or C phase from the grid to the SATs, during normal operations, operators have to diagnose the condition and manually isolate safety buses from offsite power which would automatically start and load the emergency diesel generators. During a design basis event concurrent with an open circuit on A or C phase from the grid to the SATs, analysis performed to date indicates that starting of the ECCS loads would have caused the bus voltage to decrease sufficiently to actuate the under-voltage protective relays and restore cooling with emergency onsite power without challenging fuel design limits. The 4.16kV safety bus under-voltage protection scheme at Byron and Braidwood is believed to be a typical industry design. This design issue is being evaluated at the other Exelon stations. The results of this evaluation will be shared with the NRC. Therefore, this condition is being reported as a voluntary notification due to its potential generic industry applicability." The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | ||
| ENS 47773 | 2 February 2012 02:00:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Report - Invalid System Actuation | In accordance with 10CFR50.73(a)(2)(iv)(A), this telephone notification reports an invalid actuation of the Unit 2 train B (2B) Emergency Diesel Generator (DG) on February 1, 2012, at 2100 hours. At the time of the event, Unit 2 was in Mode 5, Cold Shutdown. The 2B DG was being prepared for an operability surveillance following a planned work window. As part of this surveillance, a chart recorder is installed to monitor key DG parameters to include the DG start signal. When the second lead was connected across the starting relay contact test point, a DC ground alarm was received and the 2B DG started. The engine functioned successfully and as expected, the 2B DG did not automatically connect to its safety bus, since no bus under-voltage signal was present. The cause is attributed to a faulty chart recorder in that an inadvertent ground resulted in the actuation of the starting relay. This condition was entered into the corrective action program. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 47624 | 30 January 2012 16:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Unusual Event Due to Loss of Offsite Power Greater than 15 Minutes | At 1101 EST, Byron Unit-2 experienced a reactor trip due to RCP undervoltage. All rods fully inserted, MSIV's were manually closed and decay heat is being removed by Auxiliary Feedwater pumps running and steam leaving via atmospheric relief valves. The unit is currently in a natural circulation cooldown with the diesels supplying station emergency loads. Licensee will be cooling the plant down to Mode 5. At 1118 EST, Byron declared an Unusual Event due to a loss of offsite power on Unit 2 from a faulted Station Auxiliary Transformer (SAT). The faulted SAT caused both 6.9 kV and 4.1 kV bus voltage to drop. Smoke was observed coming from the SAT with no visible flames being apparent. This caused bus loads to trip without a complete loss of ESF busses 241 and 242. These buses were manually disconnected from the SAT, which transferred the load to the emergency diesel generators 2A and 2B. Both diesel generators started and loaded without incident. Offsite assistance was requested from the local fire department as a precaution. The licensee is also declaring notification for 10 CFR 50.72(b)(3)(v)(D) Unit 1 is not being affected by this event and remains at 100% power. The licensee has notified the NRC Resident Inspector.
At 2000 CST on 1/31/12, Byron terminated their Unusual Event due to the Loss of Offsite Power to Unit 2. Switchyard repairs were completed and offsite power has been restored to essential busses 241 and 242 thru System Auxiliary Transformers 242-1 and 242-2. Unit 2 Emergency Diesel Generators have been shutdown. The licensee is citing classification 10 CFR 50.72(c)(1)(iii) The licensee has notified the NRC Resident Inspector. Notified R3DO (James Cameron), NRR EO (Louise Lund), IRD MOC (Scott Morris), DHS (Konopka) and FEMA (Hollis). Licensee may issue a press release. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 47277 | 19 September 2011 15:07:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Spds and Erds Unavailable During Replacement of Plant Process Computer System | At 1007 CDT on September 19, 2011, the Unit 2 Plant Process Computer (PPC) was removed from service for a planned replacement in the current Unit 2 Refueling Outage. The Unit 2 PPC feeds the Safety Parameter Display System (SPDS) used in the Main Control Room (MCR) and the Technical Support Center (TSC). The Unit 2 PPC also feeds the Emergency Response Data System (ERDS). The Unit 1 and Unit 2 PPCs also feed the Plant Parameters Display System (PPDS) used in the MCR, TSC and Emergency Operations Facility (EOF). Meteorological data will remain available. The dose assessment program will remain functional as the Unit 1 PPC will be capable of providing the necessary data through PPDS to run the program. The dose assessment program is not affected by the Unit 2 PPC being out of service. As compensatory measures, a proceduralized backup method to fax or communicate via a phone circuit applicable data to the NRC, TSC, and EOF exists. There is no impact on the Emergency Notification System (ENS) or Health Physics Network (HPN) communication systems. The new Unit 2 PPC is scheduled to be functional on September 25, 2011. However, based on the Mode Unit 2 will be in, this will limit the number of points that would provide usable data. The Unit 2 PPC will be tested as Mode changes occur. The Unit 2 PPC is planned to be declared functional by Mode 2. A follow-up ENS call will be made once the Unit 2 PPC is declared functional. The loss of SPDS and ERDS is a 'major loss of assessment capability' and is reportable under 10CFR50.72(b)(3)(xiii). The NRC Senior Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS call.
As of 1800 CDT on October 8, 2011, the Unit Two PPC is considered operational with respect to the Safety Parameter Display System (SPDS), Plant Parameter Display System (PPDS) and Emergency Response Data System (ERDS). Therefore, a major loss of assessment capability no longer exists on Unit 2. The Byron EP manager contacted the NRC ERDS Center yesterday on 10/7/11 to conduct an ERDS test for Unit 2 to ensure the data was being satisfactorily sent to the NRC. Unit-1 and Unit-2 ERDS data to NRC was tested satisfactory IAW EP-AA-124-F-01. The NRC Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS update. R3DO (Phillips) notified. | Emergency Response Data System Safety Parameter Display System | |
| ENS 46708 | 30 March 2011 01:00:00 | 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 | Potential Voiding in Auxiliary Feedwater Alternate Suction Line | The design of the Auxiliary Feedwater (AF) system is for the AF pumps to normally take suction from the condensate storage tank. If the condensate storage tank is not available, the essential service water system provides the alternate supply. Due to the AF system suction piping and valve configuration, a voided section of pipe could exist in the portion that isolates the condensate storage tank supply from the essential service water supply. A preliminary vendor analysis has determined that the void fraction to reach the pump in a dynamic scenario exceeds the acceptance criteria for AF pump operability. Based on past operation in this configuration, the event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). Further review of the void model and pump performance characteristics are planned. In 2011, prior to the completion of this analysis. The void was refilled and verified full for both trains at Byron U1 and U2. Unit 1 is defueled. This condition affects both 'A' and 'B' trains of auxiliary feedwater for both Unit 1 and Unit 2. The NRC Resident Inspector has been notified. | Service water Auxiliary Feedwater | |
| ENS 46686 | 19 March 2011 13:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Ultrasonic Examination Results in Indications on Two Reactor Head Penetrations | On March 19, 2011, during the Byron Station Unit 1 refueling outage, it was determined that the results of planned ultrasonic (UT) examinations performed on two penetrations of the reactor vessel head would not meet the applicable acceptance criteria. Both require repair prior to returning the vessel head to service. These indications are not in the reactor coolant pressure boundary; however they are very near the toe of the J-groove weld. The examinations were being performed to meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to ensure the structural integrity of the reactor vessel head pressure boundary. The UT examinations continue for the remaining head penetrations. All of the penetrations will be examined during the current refueling outage. Repairs are currently being planned and will be competed prior to startup. This is reportable pursuant to 10CFR50.72(b)(3)(ii)(A) since the as found indications did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. The NRC Resident Inspector has been notified.
Ultrasonic examination (made on March 19, 2011) resulted in indications on two reactor head penetrations. As mentioned in Event Notification 46686, reactor vessel head penetrations (In-service inspection) ISI examinations were still in progress. On Wednesday, March 30, 2011, two additional Unit 1 reactor head penetrations were found to contain indications that will require repair prior to returning the reactor head to service. The indications on these two penetrations are within the reactor vessel head pressure boundary. The NRC Resident Inspector has been notified by the licensee. Notified the R3DO (Peterson). | ||
| ENS 46614 | 14 February 2011 15:20:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty Report Involving a Non-Licensed Supervisor | 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 suspended. Contact the Headquarters Operations Officer for additional details. The licensee informed the NRC Resident Inspector. |