Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56560 | 6 June 2023 13:37:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to ON-SITE Fatality | The following information was provided by the licensee via email: At 0937 EDT on June 6, 2023, it was discovered that a site employee suffered a non-work-related fatality. The individual was found non-responsive outside the Radiological Controlled Area. This is a four-hour, non-emergency notification for which a notification to other government agencies has been made. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Region I inspector has been notified. | ||
| ENS 55455 | 10 September 2021 14:55:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 72.75(b)(2), Offsite Notification | Offsite Notification Due to Contractor Fatality | This is a four-hour notification, non-emergency for a notification of another government agency. This event is being reported under 10 CFR 50.72(b)(2)(xi) and 10 CFR 72.75(b)(2). At 1055 EDT on 9/10/21, an employee of a site contractor that was performing work under a contract and in possession of the immediate area where the work was being performed, was involved in a material handling accident in the owner controlled area at Three Mile Island. Londonderry Township EMS and Fire responded to render assistance to the individual. Upon arrival to the site, medical personnel declared the individual deceased. The fatality was work related and the individual was outside of the Radiological Controlled Area. | ||
| ENS 54283 | 17 September 2019 14:28:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | On-Site Non-Work Related Fatality | Event of Public Interest performed to notify State and Local agencies for emergency vehicle response required due to an on-site non-work related illness. The individual was unresponsive and was unable to be resuscitated due to the medical issue. The individual was outside the Radiological Controlled Area (RCA) and no radioactive material or contamination was involved. The NRC Resident Inspector was notified. Responding to the site were emergency medical services, fire, and police. The licensee notified Pennsylvania Emergency Management Agency, Dauphin County Emergency Management Agency, Cumberland County Emergency Management Agency, Lancaster County Emergency Management Agency, York County Emergency Management Agency, and Lebanon County Emergency Management Agency. | ||
| ENS 53552 | 16 August 2018 04:00:00 | 10 CFR 26.719, FFD Reporting requirements | En Revision Imported Date 1/9/2019 | EN Revision Text: FITNESS-FOR-DUTY TEST POSITIVE FOR NON-LICENSED EMPLOYEE At 1519 EDT on August 16, 2018, Exelon determined a non-licensed employee had a confirmed positive for a controlled substance during a random Fitness-for-Duty test. The employee's unescorted access to the plant has been terminated. The NRC Resident Inspector has been notified.
The following is a correction to the reason for the Fitness-for-Duty test: At 1519 EDT on August 16, 2018, Exelon determined a non-licensed employee had a confirmed positive for a controlled substance during a follow-up Fitness-for-Duty test. The employee's unescorted access to the plant has been terminated. The NRC Resident Inspector has been notified of this correction. Notified R1DO (Bower) and FFD E-mail group. | ||
| ENS 53323 | 11 April 2018 09:42:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Hydrazine Spill | On April 11, 2018, a hydrazine spill resulted in measurable hydrazine levels released at the station outfall to the Susquehanna River over an approximately 16 minute time period. The hydrazine levels exceeded the station NPDES (National Pollutant Discharge Elimination System) permit effluent limitations. The Industrial Waste Treatment system release to the river was secured and no further hydrazine was released to the river. The concentrations released did not threaten the downstream users or the environment. Pennsylvania Department of Environmental Protection was notified of the NPDES non-compliance on April 11, 2018 at 0542 EDT. Pursuant to 10 CFR 50.72(b)(2)(xi), this notification satisfies the requirement to notify the NRC of the occurrence of any event or situation related to the health and safety of the public or onsite personnel, or protection of the environment, for which notification to other government agencies has been made. The NRC Resident Inspector has been notified. | ||
| ENS 53106 | 6 December 2017 14:58:00 | 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 | Diesel Fuel Oil Tank Supply Non-Conforming with Licensing Basis for Tornado Generated Missiles | On December 6, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornado generated missiles, Three Mile Island Nuclear Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. A tornado could generate a missile that could strike the emergency diesel generator (EDG) fuel oil supply tank (DFT) vent stack. This could result in crimping of the stack, which could affect the ability of the DFT to perform its 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. | Emergency Diesel Generator | 05000289/LER-2017-004 |
| ENS 52947 | 5 September 2017 15:11:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Primary Containment Declared Inoperable Due to Both Airlock Doors Open Simultaneously | On September 5, 2017 at 1111 EDT, with the reactor at 100 percent core thermal power and steady state conditions, plant personnel identified that both doors of the emergency personnel airlock of the equipment hatch were open simultaneously due to failure of the interlock. Personnel were at both the outside and inside doors, personnel heard air movement through the air lock. Immediate action was taken to close the inner containment personnel airlock door and it was verified closed. Both doors of the emergency personnel airlock of the equipment hatch were open for less than one minute. There was no radioactive release as a result of the event. The cause of the interlock failure is under investigation. This condition requires an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(ii)(A), the condition of the nuclear power plant, including its principal safety barriers (primary containment), being seriously degraded. This condition is also reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The NRC Resident Inspector has been notified. | Primary containment | |
| ENS 52927 | 25 August 2017 15:15:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Exceeding Chlorine Effluent Limit | On August 24, 2017 a malfunction internal to the River Water Chemistry Control System resulted in measurable free available chlorine levels over a one-hour period at the station outfall to the Susquehanna River. The chlorine levels exceeded the station NPDES (National Pollutant Discharge Elimination System) permit effluent limitations. The River Water Chemistry Control System was shutdown until the condition is corrected. The concentrations released did not threaten the downstream users or the environment. Pennsylvania Department of Environmental Protection was notified of the NPDES non-compliance on August 25, 2017 at 1115 (EDT). Pursuant to 10 CFR 50.72(b)(2)(xi), this notification satisfies the requirement to notify the NRC of the occurrence of any event or situation related to the health and safety of the public or onsite personnel, or protection of the environment, for which notification to the other government agencies has been made. The NRC Resident Inspector has been notified. Notified DOE, EPA, USDA, HHS, and FEMA. | ||
| ENS 52767 | 24 March 2017 18:25:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid System Actuation During Testing | On March 24, 2017, at 1425 EDT, while performing Engineered Safeguards Actuation System (ESAS) quarterly High Pressure Injection/Low Pressure Injection Logic and Component testing, an unintended test signal was generated when a test switch was moved to the OFF position but went slightly past this position and engaged contacts for the Test no. 1 position. When examined, the test switch was found to be degraded which allowed the switch to move past the center position and engage the test no. 1 contacts. This resulted In a partial actuation of 'B'- train ESAS components. It also resulted in an injection to the reactor coolant system (RCS). The test signal was immediately removed by operators and the inadvertently started equipment secured. The plant was operating at 100% power when the event occurred. There were no valid signals or plant conditions present to warrant the safety system actuation. The 'B' Emergency Diesel Generator rolled on air start but did not get up to full speed. Decay Heat Removal Pump 'B' started and the Decay Heat Removal Injection valve 4B opened, Make-Up Pump 'C' started, Make-Up Pump suction valve 14B opened, Make-Up pump discharge valves 16C and 16D opened, Spent Fuel Pump 1B tripped off, Air Handling Fan 18 tripped off and Air Handling Fan 1C trip tripped off. These components properly functioned from the inadvertent test signal and were secured prior to any adverse impact to plant operation. There was a small injection of borated water into the RCS. The plant remained stable at 100% power operation. Pursuant to 10 CFR 50.73(a)(1) the following information is provided as a sixty (60) day telephone notification to the NRC. This notification, reported under 50.73(a)(2)(iv)(A), is being provided in lieu of the submittal of a written LER to report a condition that resulted in an invalid partial actuation of the 'B' train of the Engineered Safeguards Actuation System (ESAS) as it was not part of a pre-planned sequence. The Licensee notified the NRC Resident Inspector. | Reactor Coolant System Emergency Diesel Generator Decay Heat Removal | |
| ENS 52415 | 7 December 2016 23:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Unisolable Reactor Coolant System Boundary Leakage | While TMI (Three Mile Island) Unit 1 was in a hot shutdown condition, leakage was identified coming from an RCS (reactor coolant system) pressure boundary on a welded connection on the 'A' Reactor Coolant Pump. The leakage is unisolable from the RCS and is less than 0.5 gpm. Planned actions are to cooldown Unit 1 to cold shutdown conditions in order to repair the leakage. The licensee will be notifying the Pennsylvania Emergency Management Agency and has notified the NRC Resident Inspector. | Reactor Coolant System | |
| ENS 52097 | 18 July 2016 20:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Offsite Communications Capability | At 1600 EDT, testing of the Everbridge ERO (Emergency Response Organization) notification system identified the system was not able to notify all ERO individuals. This constitutes a loss of offsite communications capability. The issue was reported resolved by the vendor and site testing has verified resolution as of 2130. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a loss of communications capability. The NRC Resident Inspector has been notified. Some of the ERO personnel did not receive a test page. The requirement is to have all ERO personnel receive the page within ten minutes. Compensatory measures were instituted while the system was not functional. | ||
| ENS 52049 | 28 June 2016 14:55:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both a & B Trains Hpi Inoperable Due to Void in Common Suction Line Resulting in Loss of Safety Function | At 1055 (EDT) on 06/28/16 a gas void was found during the monthly surveillance inspection located in the common suction line to the High Pressure Injection / Makeup (HPI / MU) pumps. At 1150 on 06/28/16 the HPI suction line cross-connect valves were closed to isolate and separate the 'A' & 'B' Trains of HPI. The 'A' train of HPI was declared degraded and initiated a 72 hour LCO (Limiting Condition of Operation) under TS (Technical Specification) 3.3.2. Investigation and analysis by Engineering determined that the size of the void did not meet the acceptance criteria for system operability. Due to the size of the void and location at time of discovery, both trains of HPI were determined to be inoperable until the suction cross connect valves were closed. This condition is reportable under 10 CFR 50.72(b)(3)(v)(D) as a Condition That Could Have Prevented Fulfillment of a Safety Function to mitigate the consequences of an accident. The void is being vented to restore a water-solid condition. The last successful surveillance was conducted on 05/31/16. The cause of the void is being investigated. The NRC Resident Inspector has been informed.
Following the 8-hour 10 CFR 50.72 notification made on 06/28/16 (EN 52049), further engineering analysis determined that the as-found void size was insufficient to cause the high pressure injection pumps to become inoperable or unable to fulfill their safety function. The cause for the void continues to be under investigation including the development of actions to prevent recurrence. Void checks are being performed at an increased frequency until cause is determined, and actions to prevent recurrence are in place. As determined through analysis, both trains of HPl were operable and available such that the safety function was never lost. Therefore, this event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified R1DO (Dimitriadis). | ||
| ENS 51695 | 2 December 2015 02:19:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Emergency Feedwater Actuation | On December 1, 2015, at 2119 EST, with Unit 1 in power operation mode, during a planned maintenance activity, an invalid Heat Sink Protection System (HSPS) actuation occurred. At the time of the event, electrical maintenance technicians were verifying a HSPS relay contact state using an electrical test meter. The contact was being verified open by recording both voltage and resistance readings across the contact. The technicians first measured voltage. No voltage was found, indicating the relay contact was open, as expected. The technicians then measured for resistance across the open contact. Test meters have lower circuit impedance when measuring resistance as opposed to voltage, which can result in electrically bridging across open contacts. When the meter was installed across the open contact to obtain the resistance reading, the HSPS actuation circuit logic was completed and the inadvertent HSPS actuation occurred. The HSPS actuation resulted in the steam driven Emergency Feedwater (EFW) pump automatically starting and control valves receiving actuation set point signals. There was no emergency feed water injection into the steam generators. At the time of the inadvertent HSPS actuation, steam generator operating levels were above the HSPS actuation setpoint. The specific train and system that actuated was the Heat Sink Protection System, Emergency Feedwater System Actuation on Loss of All Reactor Coolant Pumps (RCP) Train A. The HSPS Loss of all RCP Train A actuation was complete. The EFW valves and EFW steam driven pump started and functioned successfully. This is reported under 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of HSPS Loss of all RCP Train A and in accordance with 10 CFR 50.73(a)(1), this notification of the invalid actuation is provided in lieu of a written LER. The Licensee notified the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 51455 | 6 October 2015 01:15:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Alert Declaration Due to Fire in the Auxiliary Building | At 2115 (EDT) on 10/05/15, an Alert was declared due to a fire in the Auxiliary building affecting DC-P-1A (Decay closed cooling pump 1A) and A-Train safety equipment. The fire is out. The licensee reported the fire was extinguished at approximately 2201 EDT. The fire did not hamper operations personnel responding to the fire. Offsite fire assistance was requested. The NRC Resident Inspector will be notified. The licensee notified the York Haven Power Station, PEMA (Pennsylvania Emergency Management Agency), and the counties of Cumberland, Lebanon, Lancaster, York, and Dauphin. Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
The licensee is terminating the Alert at 0009 EDT on 10/06/15 based on the following: At 2115 (EDT) on 10/05/15, an Alert (HA3) was declared due to a fire in the Decay Closed Cooling Water Pump '1A' motor and breaker. The fire was extinguished at 2201 (EDT). Following inspection by electrical maintenance the 'P' 480V bus was re-energized at 2305 (EDT) and restoration of previously running loads is in progress. Station (TMI) is in a 72-hour LCO for repairs to the Decay Closed Cooling Water Pump '1A' (DC-P-1A). TMI (Three Mile Island) is terminating the event based upon the above information. The licensee notified the York Haven Power Station, PEMA (Pennsylvania Emergency Management Agency), and the counties of Cumberland, Lebanon, Lancaster, York, and Dauphin. The licensee will issue a press release. The licensee informed the NRC Resident Inspector. Notified R1DO (Bickett), NRR (Howe) and IRD (Gott). Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email). | ||
| ENS 50932 | 27 March 2015 17:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Seismic Monitor Not Available for Emergency Plan Assessment | Three Mile Island Station has completed a review of seismic monitor performance. The seismic monitor is currently operable however, this review identified 1 time in the past 3 years that the seismic monitor was inoperable such that emergency classification at the ALERT level could not be obtained with site instrumentation. The seismic monitor was determined to be inoperable on the following date: 1) August 7, 2012 This unplanned inoperable condition of the seismic monitor was entered into the Three Mile Island Corrective Action Program when it occurred. While Exelon procedural direction allowed the use of offsite sources to obtain seismic data when the seismic monitor is incapable of assessing emergency plan Emergency Action Levels (EALs), this was not explicitly referenced in the approved EALs. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). This report is required per 10 CFR 50.72(a)(1)(ii) as an event that occurred within 3 years of the date of discovery. The licensee has notified the NRC Resident Inspector. | ||
| ENS 50654 | 4 December 2014 16:50:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Emergency Siren Inadvertently Sounding | At 1150 (EST), the Shift Manager was notified that TMI (Three Mile Island) emergency plan siren 128 was inadvertently sounding. This siren is no longer alarming. Repairs to the siren have been completed. This notification is being made under 10 CFR 50.72(b)(2)(xi), 'News release or notification of other government agency'. The licensee has notified the NRC Resident Inspector. The licensee notified applicable state and local authorities. | ||
| ENS 50108 | 12 May 2014 14:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | During a review from industry operating experience it was identified that there are three additional unprotected DC control circuits for non safety related DC motors which are routed from the turbine building to other separate fire areas (this is in addition to the one circuit that was previously identified and submitted under event #50059). Fuses used to protect the motor power conductors appear to be inadequate to protect the control conductors. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one area can cause short circuits potentially resulting in secondary fires or cable fires in other areas where the cables are routed. The secondary fires or cable failures are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. 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 NRC Resident Inspector has been notified. | 05000289/LER-2014-001 | |
| ENS 50059 | 25 April 2014 15:13:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | A review of industry Operating Experience identified that there were unprotected DC control circuits for non safety-related DC motors which are routed from the turbine building to other separate fire areas. Fuses used to protect the motor power conductors appear to be inadequate to protect the control conductors. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one area can cause short circuits potentially resulting in secondary fires or cable failures in other fire areas where the cables are routed. The secondary fires or cable failures are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. 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 NRC Resident Inspector has been notified. | 05000289/LER-2014-001 | |
| ENS 49512 | 7 November 2013 11:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System Cold Leg Drain Line Ultrasonic Indication | On Thursday, November 7, 2013, while performing planned inspections on a 2 inch reactor coolant system drain line, TMI technicians identified an indication of a flaw on the weld internal diameter of an elbow to pipe weld on the line. This flaw is determined to not meet acceptable criteria and a repair is being developed. This condition is reportable under 10CFR50.72(b)(3)(ii). No actual impact occurred during plant operations. The NRC Senior Resident Inspector has been notified. | Reactor Coolant System | 05000289/LER-2013-001 05000289/LER-2017-002 |
| ENS 48855 | 27 March 2013 12:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Ventilation Out of Service for Maintenance | This is a non-emergency eight hour notification for a loss of emergency assessment capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the work activity affects the functionality of a emergency response facility. Planned maintenance activities are being performed today (March 27, 2013) to the Technical Support Center (TSC) HVAC. The work includes both corrective and preventive maintenance to the TSC HVAC system. This work activity is planned to be performed and completed expeditiously within about 14 hours. If an emergency condition occurs that requires activation of the TSC, plans are to utilize the TSC concurrent with this work activity as long as habitability conditions allow. Additionally, plans are in place to expedite the return of the system should an emergency condition occur. The emergency response organization duty team members will be relocated to an alternate location if required by habitability conditions in accordance with emergency implementing procedures. The licensee notified the NRC Resident Inspector.
The Technical Support Center has been returned to service. The licensee has notified the NRC Resident Inspector. Notified R1DO (Krohn). | HVAC | |
| ENS 48714 | 5 February 2013 13:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Ventilation System Out of Service | This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the work activity affects the functionality of an emergency response facility. Planned maintenance activities are being performed today (February 5, 2013) to the Technical Support Center (TSC) HVAC. The work entails performance of DOP (dioctyplthalate oil smoke test) and Halide testing. This work activity is planned to be performed and completed expeditiously within about 12 hours including establishing and removing the clearance and performing post maintenance testing. If an emergency condition occurs that requires activation of the TSC, plans are to utilize the TSC concurrent with this work activity as long as habitability conditions allow. Additionally, plans are in place to expedite the return of the system should an emergency condition occur. The Emergency Response Organization duty team members will be relocated to alternate location if required by habitability conditions in accordance with emergency implementing procedures. The licensee notified the NRC Resident Inspector.
TSC ventilation has been returned to service. Notified the R1DO (Powell) via e-mail. | HVAC | |
| ENS 48625 | 26 December 2012 09:10:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Ventilation System Out of Service | This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the work activity affects the functionality of an emergency response facility. Planned maintenance activities are being performed today (December 26, 2012) to the Technical Support Center (TSC) HVAC. The work entails removing power to the system fan and dampers to perform required preventative maintenance (PM) rendering the TSC HVAC non-functional during the performance of this work activity. This work activity is planned to be performed and completed expeditiously within about 16 hours including establishing and removing the clearance and performing post maintenance testing. If an emergency condition occurs that requires activation of the TSC, plans are to utilize the TSC concurrent with this work activity as long as habitability conditions allow. Additionally, plans are in place to expedite the return of the system should an emergency condition occur. The Emergency Response Organization duty team members will be relocated to alternate locations if required by habitability conditions in accordance with emergency implementing procedures. The licensee notified the NRC Resident Inspector. | HVAC | |
| ENS 48325 | 20 September 2012 18:16:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Reactor Protection System Actuation | On September 20th at 1416 EDT, Three Mile Island automatically tripped due to a flux to flow imbalance as a result of a trip of the 'C' reactor coolant pump. The cause of the trip of the 'C' reactor coolant pump is still under investigation. The electrical grid is stable and unit 1 is being supplied by offsite power. All control rods have fully inserted. Decay heat is being removed by main feedwater flow to both steam generators that are exhausting via the normal main condenser cooling loop under manual control. Preliminary evaluation indicates that all plant systems functioned normally following the reactor trip, except for automatic operation of turbine bypass valve control due to failure of the automatic control function to control precisely at setpoint. Three Mile Island remains stable in hot shutdown mode while conducting the post trip review. No radioactive releases were experienced as a result of this event. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation, and under 10 CFR 50.72 (b)(2)(xi) due to an information release to local officials. Both are four hour reports. The licensee notified the NRC Resident Inspector." The licensee has notified the state and local governments, and will be making a media release. | Steam Generator Feedwater Reactor Protection System Main Condenser Control Rod | |
| ENS 48221 | 22 August 2012 12:01:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 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 High Reactor Coolant System Pressure | On August 22, 2012, during a planned load reduction, Three Mile Island Unit 1 Reactor automatically tripped at 0801 EDT due to high reactor coolant system pressure as a result of a main feedwater transient. The cause of the main feedwater transient is still under investigation. The Emergency Feedwater System (EFW) actuated at 0801 EDT. The electrical grid is stable and Unit 1 is being supplied by offsite power. All control rods fully inserted. Decay heat is being removed via the Emergency Feedwater System (EFW) flow to both Steam Generators that are exhausting via the normal main condenser cooling loop under manual control. Preliminary evaluation indicates all plant systems functioned normally following the reactor trip and EFW actuation except for manual operation of Turbine Bypass Valve control due to failure of the automatic control function. Three Mile Island Unit 1 remains stable in Hot Shutdown mode while conducting the post trip review. No radioactive releases were experienced as a result of this event. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation, and under 10 CFR 50.72(b)(2)(xi), due to an information release to local officials, both are four (4) hour reports. This event is also reportable under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(3)(iv)(B), due to a valid actuation of the Emergency Feedwater System. The licensee notified the NRC Resident Inspector." Licensee notified the State, local and other Government agencies. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Main Condenser Control Rod | 05000289/LER-2012-004 |
| ENS 48220 | 22 August 2012 06:28:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Tech Spec Required Shutdown Due to Reactor Coolant System Pressure Boundary Leak | On August 22, 2012, it was determined, based on a remote video camera inspection inside the secondary shield wall, that a primary system leak at a Pressurizer Heater bundle exists and is not isolable. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.1.6.4 does not allow any leakage through a nonisolable RCS strength boundary. The discovery of leakage from the Pressurizer Heater bundle is considered a degradation of a principal safety barrier. This condition does not represent a reduction in the public health and safety. This is reportable as a 4-hour ENS notification under 10CFR50.72(b)(2)(i) due to the initiation of a nuclear plant shutdown required by the plant's Technical Specifications. The licensee notified the NRC Resident Inspector. The licensee expects to also notify state and local government agencies of the shutdown. The current leak rate, as calculated by the plant process computer, is approximately .2 gallons per minute. | Reactor Coolant System | 05000289/LER-2012-003 |
| ENS 48179 | 10 August 2012 16:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Flood Barrier Seals Could Not Be Verified Installed | On August 10, 2012 during an inspection in the Air Intake Tunnel, six 4-inch conduits that carry cabling from yard vaults through the Air Intake Tunnel (AIT) to the Auxiliary Building (AB) were inspected for flood seals. This was done by opening the conduit seals bottom drain openings to inspect the condition by boroscope. These seal components are just inside the AIT from the electrical vaults. During the inspections no sealant could be readily identified. Each conduit from the yard vaults that is not sealed could potentially provide a leak path during flood conditions from the yard vaults to the Auxiliary Building via the electrical conduit. Flood water entering the Auxiliary Building could impact the decay heat removal function. This is reportable as an 8 hour ENS notification under 10CFR50.72(b)(3)(v)(B) and 10CFR50.72(a)(1)(ii) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. The licensee notified the NRC Resident Inspector. | Decay Heat Removal | 05000289/LER-2012-002 |
| ENS 48132 | 25 July 2012 18:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Press Release Due to an Elevated Level of Tritium Detected on Site | TMI is issuing a press release and performing courtesy stakeholder communications as a result of an elevated level of tritium detected in one of 55 on-site ground water monitoring wells near the plant structure. The tritium concentration in the well is contained to the site and well (within) EPA drinking water standards. TMI is performing voluntary communications in accordance with NEI 07-07 Industry Ground Water Protection Initiative and contacting local and state stakeholders as a courtesy. As a result of the press release, a four-hour notification is being made per 10 CFR 50.72 (b)(2)(xi). The licensee notified the NRC Resident Inspector, the Commonwealth of Pennsylvania and local counties (Dauphin and York). | ||
| ENS 47419 | 7 November 2011 01:03:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Spds and Plant Process Computer Due to Planned Maintenance | At 1203 EST on 11/06/2011, the Safety Parameter Display System (SPDS) was removed from service for planned maintenance of the Plant Process Computer. The Plant Process Computer and the SPDS function were restored at 2015 EST on 11/06/2011. This event is reportable per 10 CFR 50.72(b)(3)(xiii) since the SPDS was out of service for greater than 8 hours resulting in a major loss of emergency assessment capability. The licensee will inform the NRC Resident Inspector. | Safety Parameter Display System | |
| ENS 47390 | 29 October 2011 23:18:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Greater than 25% of Emergency Notification System Sirens Due to Weather Conditions | At approximately 1818 EDT on October 29, 2011, Three Mile Island Nuclear Generating Station received notification that 25 of 96 emergency notification system offsite sirens were not functional. The loss of >25% of the sirens is considered a Major Loss of Emergency Preparedness Capabilities (10CFR50.72(b)(3)(xiii)). The loss of offsite sirens was due to loss of electrical power caused by significant heavy snowfall in the area. Efforts are in place to restore offsite sirens to service. The licensee stated that compensatory measures are in place should their be a need to notify the areas affected by the lost sirens. The licensee has notified State and local authorities. The NRC Resident Inspector will be notified. | ||
| ENS 47294 | 26 September 2011 19:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | New River Hydraulic Analysis Raises Maximum Flood Level | TMI new river analysis indicates level above existing UFSAR (Updated Final Safety Analysis Report) flood analysis. At about 15:00 (EDT) on September 26, 2011, a revised River Stage Discharge Analysis was completed and concluded that the Probable Maximum Flood (PMF) water level is higher than previously described in the safety analysis report. This unanalyzed condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). Actions to protect the plant for the higher PMF river level have been implemented. The flood barrier gates that protect safety related equipment have been modified to accommodate the revised river levels. No onsite flood water levels have occurred that could potentially challenge the existing flood barrier system. The licensee notified the NRC Resident Inspector. The licensee will be making a courtesy media notification. | ||
| ENS 47190 | 23 August 2011 18:01:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of an Unusual Event Due to a Seismic Event | At 1401 hrs. EDT on 8/23/11, TMI (Three Mile Island) declared an Unusual Event due to a threshold seismic condition (HU5) earthquake. The earthquake was felt at the plant. No equipment damage has been identified. No personnel injuries were reported. The licensee has notified the NRC Resident Inspector and state and local authorities.
The Unusual Event declared at 1401 EDT due to the ground motion felt at the site has been terminated at 1730 EDT. Walkdowns have been performed and no equipment damage has been identified. The plant remains in a stable condition at 100% power. Offsite power and diesel generators were verified to be unaffected by the event. The licensee has notified the NRC Resident Inspector and state and local authorities. Notified R1 IRC (Dentil), IRD (Gott), DHS (Bean), FEMA (Via), USDA (Kraus), and DOE (Turner). | ||
| ENS 47048 | 14 July 2011 17:19:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Inadvertant Siren Actuation | At about 1319 on July 14, 2011, Dauphin County 911 (Pennsylvania) was advised by a local resident that siren number 121 sounded for 3 minutes. Dauphin County (Pennsylvania) contacted the TMI main control room to inform Exelon of an inadvertent siren actuation at about 1350 on July 14, 2011. Pennsylvania Emergency Management Agency (PEMA) later notified (about 1430 on July 14, 2011) the TMI main control room of the same issue. This notification is being made in accordance with 10CFR50.72(b)(2)(xi) because offsite agencies were notified regarding the inadvertent actuation of part of the public notification system. Siren Maintenance was notified and is troubleshooting the issue. The licensee notified the NRC Resident Inspector. | ||
| ENS 47031 | 6 July 2011 18:40:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center (Tsc) Hvac Out of Service Due to Loss of Power | At about 1440 EDT on July 6, 2011, the Technical Support Center (TSC) HVAC lost power during motor control switch testing. The transfer switch did not fully engage resulting in loss of power to the TSC HVAC fan motor. Troubleshooting was performed and power was restored to the TSC HVAC system at approximately 1625 EDT. This involved an emergent loss of the TSC capability that could not be readily remediated in less than the TSC staffing time of one hour. Unrelated to the first event, 13.2 KV power to the TMI station office buildings was lost at approximately 1740 EDT on July 6, 2011 due to a broken electric service pole. This resulted in another loss of the TSC HVAC that could not be readily remediated. At 2115 EDT power was restored and the ventilation was started. This event resulted in another reportable condition. These events affected the ability of the TSC ventilation to maintain adequate radiological habitability in the event of an emergency with an airborne radiological release. All other capabilities were unaffected by this emergent condition. Existing procedures provide direction to relocate TSC personnel in the event of a TSC habitability concern, however the backup facility does not have standby electrical power or a filtered ventilation system. The licensee has informed the NRC Resident Inspector. | HVAC | |
| ENS 46194 | 21 August 2010 15:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Flood Barriers Needed to Protect Safety Related Equipment Missing | On August 21, 2010 an inspection of the Air Intake Tunnel (AIT) sump identified missing flood barriers needed to protect safety related equipment in the plant. If enough flood water had entered into the AIT, water could have entered into the Auxiliary Building (AB) through the ventilation ductwork that connects the AIT and the AB. If flood water continued to enter the AB, then safety related equipment in the AB could have been affected. This condition could have resulted in the unavailability of equipment in the Auxiliary Building including the 1A and 1B Decay Heat pumps, the 1A and 1B Building Spray pumps and 1A, 1B and 1C Make-up Pumps. This is reportable as an 8 hour ENS notification under 10CFR50.72(b)(3)(v)(B) and 10CFR50.72(a)(1)(ii) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. Flood protection barriers have been established for the affected penetrations. Inspections of the flood protection barriers are ongoing. Further engineering review is being performed to determine the impact of the potential water intrusion into the AIT and AB. The licensee notified the NRC Resident Inspector. The licensee will notify the Pennsylvania Bureau of Radiation Protection.
EN# 46194 made on August 21, 2010 at 1100 EDT is being retracted. Upon further Engineering review it was determined that the external flood barrier design deficiency in the Air Intake Tunnel (AIT) did not result in a condition that could have prevented the fulfillment of the safety function needed to remove residual heat. The AIT sump pump is in the cavity that had the degraded barrier where the in leakage of concern enters through conduits from yard electrical vaults. This pump starts automatically and has capacity in excess of the in leakage during the Probable Maximum Flood (PMF). The AIT sump pump's functional capability was confirmed to be available over the past three years. Therefore the ability to prevent flood water from the AIT reaching and impacting safety related equipment needed to remove residual heat was maintained. The event is therefore NOT reportable. The Licensee has notified the Resident Inspector. Notified R1DO. | ||
| ENS 45977 | 4 June 2010 14:20:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - 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 terminated. Contact the Headquarters Operations Officer for additional details. | ||
| ENS 45929 | 14 May 2010 19:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Nonfunctional | At 1530 on May 14th, 2010, Three Mile Islands Technical Support Center (TSC) ventilation, filtration and climate control system was identified as nonfunctional. Site emergency implementation procedures provide direction for performance of TSC functions in alternate locations. This failure affects the ability of the TSC ventilation 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. This condition is considered a major loss of emergency assessment capability and is reportable under 10CFR50.72(b)(3)(xiii). The NRC Resident Inspector has been notified.
The TSC ventilation system has been repaired and is fully functional as of 1100 on May 15, 2010. The NRC Resident Inspector has been notified. Notified R1DO(Miller). | ||
| ENS 45514 | 22 November 2009 03:45:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Radiation Release in Containment Associated with Steam Generator Removal | This event is being reported via the ENS to the NRC Operations Center within four hours after notifications were made to the Pennsylvania Emergency Management Agency and local counties for an event of potential public interest. This notification was made at 22:45 on November 21, 2009, after the occurrence of an event related to the safety and health of onsite personnel for which a press release has been made. The press release was made following the event of potential public interest notifications. This report is being made under 10 CFR 72.75(b)(2). At approximately 16:00 on Saturday November 21, 2009, low levels of radiation activity were measured on radiation monitors installed in the TMI-1 reactor building. Personnel were directed to immediately leave the reactor building until the source of the activity could be identified. Surveys directly outside of the reactor building construction opening indicated a slight increase in activity. Levels have returned to normal. No contamination was identified outside of the reactor building. Approximately 150 workers were monitored for exposure to the radiation activity. No worker approached or exceeded any exposure limits. The sources of the activity are believed to be from maintenance tasks related to cutting lines in preparation for removal of the 'B' Steam Generator. The licensee is still investigating the cause of the event but indicated that the radiation release is no longer in progress. Containment Ventilation is established to provide an inflow into containment to the extent possible (considering the openings in containment). The licensee contacted state and local authorities and the Pennsylvania Emergency Management Agency (PEMA). The licensee notified the NRC Resident Inspector. The licensee has also issued a press release.
This provides an update to notification EN #45514 that reported a non-emergency offsite notification for an event on November 21, 2009, at the Three Mile island Nuclear Station. Updated news release and notifications to local, state and federal elected and regulatory officials were performed today, December 18, 2009, that provide an outage update as well as communication the results of analysis of Offsite Environmental Monitors following the November 21, 2009 event. The licensee notified NRC Region I. Notified R1DO (Barkley). | Steam Generator | |
| ENS 45464 | 23 October 2009 21:41:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Steam Generator Safety Relief Valves Did Not Open at Required Setpoint | Online testing of the secondary side pressure relief valves revealed that 3 valves did not open at the required pressure set point on the initial tests. All valves did open on subsequent tests at a higher pressure and have been restored to operable status. These 3 valves and 3 other similar valves were installed during the last refueling outage and it is concluded that all 6 valves have the same set point drift failure mechanism. All 6 valves are assumed to have been inoperable at some time during the last operating cycle. The steam generator design analysis specifies a minimum pressure relief capability to prevent a failure of the steam generator and subsequent potential loss of decay heat removal capability. This condition could have prevented the fulfillment of the safety function needed to remove residual heat. An engineering analysis is in progress to determine the actual impact on the safety function. Per Technical Specifications 3.4.1.2.3, reactor power was reduced to below 66.3% due to the remaining 3 inoperable relief valves. The licensee notified the NRC Resident Inspector.
In addition to the 6 valves discussed in EN 45464, an additional valve was found to be inoperable. Thus, it is assumed that 7 of the 18 secondary side pressure relief valves were inoperable at some time during the last operating cycle. An engineering analysis has determined that the inoperable secondary side pressure relief valves did not result in a condition that could have prevented the fulfillment of the safety function needed to remove residual heat. The impact of the inoperable secondary side pressure relief valves on the steam generators and associated secondary side piping integrity was assessed by comparison to an existing analysis. A qualitative assessment of that analysis indicates that with the reduced steam relief capacity, the pressure in the steam generators and associated secondary side piping would remain below the ASME code allowable. Thus, there would have been no failure of the steam generators and no subsequent potential loss of decay heat removal capability. Therefore, this event does not meet the criteria of 50.72(b)(3)(v)(B) and is being retracted. This event will be reported as a 60 day LER in accordance with 10CFR50.73(a)(2)(i)(B). The licensee notified the NRC Resident Inspector. Notified R1DO (Holody). | Steam Generator Decay Heat Removal Safety Relief Valve | |
| ENS 45359 | 17 September 2009 17:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Emergency Response Data System Inoperable for Greater than Eight Hours | At 1300 hrs. on Thursday, September 17, 2009, TMI Unit 1 determined that there had been a degradation of the emergency preparedness response capabilities when a loss of the Emergency Response Data System (ERDS) was identified. While preparing for a future job in the same cabinet where the ERDS modem is located, technicians identified that the phone line for the ERDS modem was not connected. At 1400, the phone line was reconnected to the modem. After coordination with the NRC Operations Center testing personnel, a test of the ERDS system was initiated. At 1414, an active link was established. A successful ERDS link was confirmed with the NRC Operations Center. At 1426, the ERDS link was terminated. Work was performed on different equipment in the cabinet containing the ERDS modem on approximately September 1, 2009. It is probable that the phone line was inadvertently disconnected from the ERDS modem at that time. The last scheduled quarterly test was successfully completed on July 9, 2009. TMI 1 has determined that this event is reportable to the NRC as an 8-hour non-emergency report in accordance with 10 CFR 50.72 (b)(3)(xiii). The licensee has notified the NRC Resident Inspector. | Emergency Response Data System | |
| ENS 45231 | 27 July 2009 14:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center (Tsc) Hvac Found Degraded Due to Water Intrusion | At about 10:00, on July 27, 2009, the Technical Support Center (TSC) HVAC was found to be degraded. The fan motor was running but there was no air flow and the TSC rooms were not being maintained with a positive pressure. Upon investigation, the fan housing was found partially filled with water, submerging the motor and preventing air flow through the system. Repair of the motor will take more than one day and is being immediately pursued. This affects the ability of the TSC ventilation 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 repair. Existing procedures provide direction to relocate TSC personnel in the event of a TSC habitability concern; however, the backup facility does not have standby electrical power or a filtered ventilation system. Therefore, this condition is considered a major loss of emergency assessment capability and is reportable under 10CFR50.72(b)(3)(xiii). The TSC HVAC was last functionally tested satisfactorily on 06/29/09. The licensee believes the water found in the HVAV fan housing may be a result of a clogged drain line. The licensee will inform the NRC Resident Inspector. | HVAC | |
| ENS 45324 | 10 July 2009 06:38:00 | 10 CFR 50.73(a)(1), Submit an LER | Electrical Short Circuit Resulted in Partial High Pressure Injection | This event is being reported via a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. In this case, the telephone report is not considered an LER. This report is being made under 10 CFR 50.73(a)(2)(iv)(A). During a maintenance activity to replace an Engineered Safeguards Actuation System (ESAS) relay, one of the two High Pressure Injection (HPI) valves (MU-V-16C) in the 'B' train partially opened. The HPI system consists of the 'A' and 'B' trains, with each train containing two HPI valves. The cause of the partial opening of the HPI valve was inadvertent contact with adjacent energized circuits during the replacement of the ESAS relay. This resulted in a momentary short circuit, which bypassed the normal actuation logic, causing the HPI valve to open approximately 0.17 inches before blowing the control power fuse that stopped the valve movement. The event resulted in the inadvertent transfer of approximately 1000 gallons of water from the Make-up Tank into the Reactor Coolant System (RCS), before the valve could be restored to the closed position. No other valves or components actuated as a result of the inadvertent short circuit. RCS volume and pressure were stabilized and returned to normal. The 'B' HPI train had been declared inoperable and the unit entered a 72 hour LCO at 1:03 AM on 7/10/2009 due to configuration requirements needed for the planned ESAS relay replacement. The inadvertent partial 'B' train HPI did not impact the 'A' HPI train, and the unit remained at full power during this event. Following troubleshooting and replacement of the control power fuse, MU-V-16C was tested and restored to OPERABLE at 4:59 AM on 7/11/2009. The licensee notified the NRC Resident Inspector. | Reactor Coolant System | |
| ENS 44715 | 24 October 2008 18:18:00 | 10 CFR 50.73(a)(1), Submit an LER | Inadvertent Actuation of Engineered Safeguards System from Relay Sensing Circuit | This event is being reported via a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. In this case, the telephone report is not considered an LER. This report is being made under 10CFR 50.73 (a)(2)(iv)(A). During Engineered Safeguards Actuation System (ESAS) logic testing on October 24, 2008, an invalid actuation of the following heat removal systems occurred: 'B' train of the Decay Heat River Water System (DR), 'B' train of the Decay Heat Closed Cooling Water System (DCCW), and the 'B' train of the Nuclear Services River Water System (NR). There was no injection into the Reactor Coolant System. The invalid actuation occurred when the channel under test was taken to its tripped position. Since ESAS utilizes a 2 out of 3 logic for actuation, another actuation signal was present on one of the two channels not being tested, satisfying the actuation logic for the affected systems. The invalid actuation of these heat removal systems during testing on October 24, 2008 was due to oxidation on a silver-plated contact in one of the other two channels that was not being tested. This contact oxidation caused a higher input resistance to the timer relay, which resulted in an inadvertent actuation of the relay and its associated systems. The contact oxidation was caused as a result of using silver plated contacts in a low voltage application (approximately 12 VDC). During this invalid actuation, the heat removal systems were fully actuated. These heat removal systems functioned successfully and the operation of these systems did not have any adverse impact on plant operation. All of the silver-plated contacts in the affected circuits will be replaced with gold-plated contacts. The contacts are scheduled to be replaced by December 18, 2008. The NRC Resident Inspector has been notified. | Reactor Coolant System | |
| ENS 44295 | 13 June 2008 15:00:00 | 10 CFR 21.21, Notification of failure to comply or existence of a defect and its evaluation | Part 21 Report Involving Commercial Grade Relay Contacts Containing Potential Defects | The following Part 21 notification was received via fax: On June 13, 2008, AmerGen Energy Company, LLC (AmerGen) completed a reportability determination which concluded that relay contacts contained an underlying design vulnerability that created a failure mode, and were reportable under Part 21, since the underlying vulnerability could create a substantial safety hazard. The relay contacts are Commercial Grade items dedicated by AmerGen. The relay contacts are provided by Joslyn Clark Controls Inc. (formerly AO Smith) as Normally Closed (N/C) open top contact assemblies (part numbers KPM-44, KPM-46, KPM-6A, and KPM-4A). These contacts are used in safety related applications, primarily in the engineered safeguards actuation system (ESAS). They are also used in safety related applications in the makeup/high pressure injection (HPI) system, main steam system, and the heat sink protection system (HSPS). The underlying vulnerability associated with the N/C Joslyn Clark contact is the design of the nylon contact arm. The design allows the contact to become configured incorrectly during assembly or following maintenance. If installed improperly, the moving contactor can hang up on the lip of the slot in the nylon actuator. The hang up results in failure of the contact to fully close and perform its function. Following proper assembly, the N/C Joslyn Clark contacts will not become hung-up during normal relay operation. As a result of this exposed design vulnerability, the Commercial Grade Dedication (CGD) plans as well as the maintenance procedures have been enhanced to prevent a relay from being placed into service with an improperly configured contact. Extent of condition reviews performed to date on relays installed in the plant, have not identified any of these deficiencies. The NRC resident inspector was notified of this part 21 notification by the licensee. | Main Steam | |
| ENS 43780 | 13 November 2007 11:32:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Unexpected Reactor Vessel Water Level Change During a Refueling Outage with Irradiated Fuel in the Core | While in a drained down condition the reactor vessel experienced an unexplained loss of reactor coolant system inventory with irradiated fuel within the vessel. Licensee entered Emergency Action Level MU-9, Unusual Event at 0632 EST. Within a few minutes after entering EAL MU-9 reactor vessel water level returned to its initial water level of 13.7 inches above the centerline of the Hot Legs of the Reactor vessel. Licensee is investigating why reactor vessel water level changed. Chairman Klein's Comm. Assist. (Bill Orders), Comm. Jaczko's Assist (T. Hipschman), Comm. Lyons Assist (S. Baggett), R1DO (M. Miller), R2DO (M. Lesser), R3DO (M. Phillips), R4DO (V. Campbell) , DHS (A. Ackers), FEMA (Sullivan), DOE (S. Bailey), USDA (Watts), HHS (C. Harper) , EPA (NRC) Chief Brown were notified.
Termination of Unusual Event. EAL classification of MU9 is no longer applicable. Investigation of the unplanned Reactor vessel level showed no indication of leakage into any building containing RCS piping. The apparent cause of the event is due to a slight negative pressure in the RCS. When the OTSG lower manways were installed with HEPA filters fan units running, the indicated RCS level lowered and stabilized due to the HEPA units drawing air from the RCS. No actual RCS inventory loss occurred. When the HEPA units were turned off the RCS level returned to the indicated Reactor vessel level prior to the event. In accordance with station procedures this EAL event was terminated at 0929 on 11/13/07. The licensee notified the Resident Inspector, State and Local Officials of this event update R1DO (D. Holody), NRR EO (J. Dozier), IRD (B. McDermott) , DHS ASWO (Ed Hoisinton) and FEMA (D. Sullivan) were notified of the termination of the Unusual Event. | Reactor Coolant System | |
| ENS 43772 | 5 November 2007 21:49:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Spds Out of Service for Planned Maintenance | At 16:49 hours, on 11/05 the Unit 1 SPDS and ERDS systems were removed from service to support restoration activities from a planned maintenance outage on the power supply. The duration of work is expected to be approximately 10 hours. (Scheduled for completion at 03:00 on 11/06/2007). During this time, Control Room indications and alternate methods will be available. Since the SPDS computer system will be unavailable for greater than 8 hours, this is considered a Loss of Emergency Assessment Capability and reportable under 10 CFR 50.72(b)(3)(xiii). A follow-up notification will be made to the NRC after the SPDS and ERDS system are returned to service. The licensee will notify the NRC Resident Inspector. * * * UPDATE AT 0315 ON 11/6/07 FROM J. PAULES TO P. SNYDER * * * At 0314 on 11/6/07 the SPDS and ERDS systems were returned to service. The licensee will notify the NRC Resident Inspector. Notified R1DO (M. Miller). | ||
| ENS 43760 | 31 October 2007 03:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Outage of Safety Parameter Display System (Spds) and Emergency Response Data | SYSTEM (ERDS) At 2330 hours, on 10/30/2007, the SPDS and ERDS system was removed from service to support activities from a planned maintenance outage on the 'C' vital bus power supply. The duration of work is expected to be approximately 12 hours. (Scheduled for completion at 11:30 hours on 10/31/2007). During this time, Control Room indications and alternate methods will be available. Since the SPDS computer system will be unavailable for greater than 8 hours, this is considered a Loss of Emergency Assessment Capability and reportable under 10CFR50.72(b)(3)(xiii). The licensee will notify the NRC Resident Inspector. Notified R1DO (Dental).
At 0230 hours, on 11/1/2007, the SPDS and ERDS system were returned to service. The licensee will notify the NRC Resident Inspector. Notified R1DO (Dental). | Safety Parameter Display System | |
| ENS 43498 | 27 June 2007 12:28:00 | 10 CFR 50.73(a)(1), Submit an LER | Failed Relay Starts One Train of the Decay Heat River Water System | This event is being reported via a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. In this case, the telephone report is not considered an LER. This report is being made under 10CFR 50.73 (a)(2)(iv)(A). During Engineered Safeguards Actuation System (ESAS) logic testing on June 27, 2007, an invalid actuation of the 'A' train of the Decay Heat River Water System (DR) occurred. The DR is comprised of the 'A' and 'B' trains, and serves as the reactor's ultimate heat sink. The ESAS has three independent input channels, which cause actuation in a two out of three logic. During ESAS logic testing, the channel under test is placed in the tripped state. Actuation relays in the other two channels should remain energized by the outputs from the other two channels not under test. The invalid actuation of the 'A' DR train during testing on June 27, 2007, was due to a failed ESAS relay in one of the other two channels not being tested. During this invalid actuation, the 'A' DR train was fully actuated. The operation of this normally standby system did not have any adverse impact on plant operation, nor any negative impact on the DR. The failed relay that caused the invalid actuation during ESAS logic testing did not impact the OPERABILITY of the ESAS, since the relay was failed in the actuated state. The failed relay was replaced on June 29, 2007. The licensee notified the NRC Resident Inspector. | ||
| ENS 43139 | 2 February 2007 18:50:00 | 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 | Low Pressure Injection (Lpi) Net Positive Suction Head Flow Requirements Not Met for Certain Accident Sequences | At 1350 hours on February 2, 2007, with the plant at 100% power, it was determined (that) the low pressure injection (LPI) system net positive suction head calculation does not account for the additional flow through the failed LPI pump recirculation line during certain accident scenarios. The additional flow is upstream of the flow element used by control room operators to throttle system flow to maintain net positive suction head flow requirements. The additional flow could result in net positive suction head below required design limits. The system design is not affected in events where both LPI trains perform as designed. Emergency operating procedures direct control room operators to open the LPI system discharge flow cross-connect line isolation valves, if accessible, following a LPI pump failure. Operators are then directed to throttle system flow through the operable LPI pump to maintain proceduralized values. These values are designed to provide sufficient design flow and maintain pump NPSH. During a simulator training scenario, operators identified when the discharge cross-connect line isolation valves were opened, the idle Building Spray train indicated flow. Follow-up investigation identified the increased flow was due to back flow through the failed LPI pump minimum flow recirculation line. This additional flow is upstream of the flow element used by operators to maintain adequate net positive suction head for the operable LPI pump. The additional flow could result in not meeting NPSH design requirements. The licensee entered the 72 hour Technical Specification limiting condition for operation (LCO) for one inoperable LPI train. The licensee is revising calculations and emergency operating procedures to account for the additional flow. This condition is reportable in accordance with 10CFR 50.72(b)(3)(ii) and (b)(3)(v) as an unanalyzed condition, and a condition that could have prevented the fulfillment of the safety function of the LPI system to mitigate the consequences of an accident, respectively. The NRC Resident Inspector was notified of this event by the licensee.
The purpose of this report is to retract the ENS report made on February 2, 2007 at 2105 hours ( ENS #43139) under 10CFR50.72(b)(3)(ii) and (b)(3)(v) as an unanalyzed condition, and a condition that could have prevented the fulfillment of the safety function of the Low Pressure Injection (LPI) system to mitigate the consequences of an accident, respectively. The initial report was made when it was determined that the LPI system net positive suction head (NPSH) calculation does not account for the additional flow through the LPI pump recirculation line during certain accident scenarios. The additional flow could result in net positive suction head below required design limits. Due to this condition, it was not certain if the LPI system could have met its design basis requirements. The licensee entered the 72 hour Technical Specification limiting condition for operation (LCO) for one inoperable LPI train. The LCO was exited on February 3, 2007 at 9:25PM following implementation of a procedure change that accounted for the additional flow and ensured that adequate NPSH was maintained. A subsequent engineering evaluation has determined that sufficient LPI pump NPSH would have been available to perform its design basis function prior to the procedure change. The engineering evaluation shows that the LPI pumps remained capable of performing their design basis functions based on the following three independent assessments: 1) the LPI pumps would have operated well beyond their mission time without significant cavitation damage at the available NPSH 2) proceduralized operator actions would have throttled flow to restore required NPSH if signs of cavitation occurred 3) an evaluation using realistic Reactor Building pressures showed that sufficient NPSH would exist. The licensee notified the NRC Resident Inspector. Notified R1DO (Hott) | ||
| ENS 43050 | 13 December 2006 22:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit Reactor Trip Due to Grid Disturbance | While operating at 100 % power the unit experienced an automatic Reactor trip from Reactor Protection System actuation at 17:48 on 12/13/06. At the time of the trip there was a grid disturbance followed by the unit tripping. The cause of the trip is under investigation. All plant systems functioned as designed. The unit is being maintained in Hot Shutdown conditions using applicable plant procedures. All control rods inserted fully after the reactor trip. S/G safety valves lifted and reset per design during the transient. The S/Gs are being supplied with normal feedwater and decay heat is being removed to the main condenser via the steam dump valves. The electric plant is in a normal shutdown lineup and the EDGs are available. The licensee notified the NRC Resident Inspector, PEMA, the Counties of Dauphin, Cumberland, Lancaster, York, and Lebanon. | Feedwater Reactor Protection System Main Condenser Control Rod | 05000289/LER-2006-003 |
| ENS 42957 | 2 November 2006 18:34:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Turbine Trip/ Reactor Trip Due to Invalid Low Condenser Vacuum Signal | At 1334 on 11-2-06 an Automatic Reactor Trip occurred from 100% power. All systems functioned as required. One safety valve stuck open on both OTSGs. They subsequently re-seated. An employee working on the roof at the time of the trip fell off a ladder and injured his leg. Emergency medical was contacted to assist with the injured worker. Two fire trucks and an ambulance was dispatched to the site to remove the injured worker. The worker was not contaminated. There is no indication of any OTSG tube leaks. Initial investigation indicates the reactor tripped, due to a turbine trip due to an invalid low vacuum signal. State and local officials will be notified of this event by the licensee. I&C Techs were performing maintenance on one of the low pressure vacuum switches. An electrical fault fed to the other two low pressure vacuum switches causing a 2 out of 3 signal which resulted in a turbine trip followed by a reactor trip signal, as expected. All rods fully inserted into the core. One safety valve (9 safety valves on each OTSG) on each Once Through Steam Generator stuck open. OTSG "B" safety relief valve was open less than one minute. There are no leaking OTSG tubes. A condensate relief valve located in the turbine building opened/shut - nobody injured. The ICS (Integrated Control System) operated as expected. All emergency core cooling systems and the emergency diesel generators are fully operable plus the electrical grid is stable. A licensee working on the industrial coolers on top of the industrial building, standing on a ladder, fell off the ladder when OTSG relief valve opened. Licensee either broke or badly sprained his leg. The NRC Resident Inspector was informed of this event by the licensee.
Post trip evaluation determined that the Main Steam safety valves were not stuck open. The safety valves were operating within their tolerance band. The "B" OTSG Main Steam safety valve reseated with no operator action as steam pressure decreased. The "A" Main Steam safety valve was reseated when operators lowered OTSG pressure in accordance with Plant Operating Procedures. TMI-1 issued a press release on this event at 15:13 on 11/2/06." R1DO (John White) notified. The NRC Resident Inspector was notified of this update by the licensee. | Steam Generator Emergency Diesel Generator Main Steam Safety Valve Emergency Core Cooling System Safety Relief Valve | 05000289/LER-2006-002 |