Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 57003 | 3 March 2024 17:42:00 | 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 Loss of Main Feedwater Pump Suction | The following information was provided by the licensee via email: At 1142 CST on 3/3/2024, with Unit 2 in Mode 1 at 29 percent power, the reactor automatically tripped due to a turbine trip caused by a loss of suction to the 22 main feedwater pump. All systems responded normally post trip. Decay heat is being removed via the auxiliary feedwater water system. Secondary steam control mechanism is the steam generator PORVs (power operated relief valves). Unit 1 remains at 100 percent power and is unaffected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The resident NRC inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The trip occurred while the licensee was returning to power operations after a refueling outage. During the trip, all rods inserted into the core. The plant is in a normal shutdown electrical lineup with offsite power available. The plant will be maintained at normal operating temperature and pressure. There is no known primary to secondary leakage. The cause of the loss of 22 main feedwater pump suction is under investigation. | |
| ENS 56827 | 1 November 2023 21:52:00 | 10 CFR 26.719, FFD Reporting requirements | Controlled Substance Found in Protected Area | The following information was provided by the licensee via email: On October 31 at 1856 CDT, Prairie Island Nuclear Generating Plant personnel identified a prohibited item (alcohol) in a kitchen area located within the protected area. An 'Extent of Condition' search was performed of all other protected area kitchen areas, no additional prohibited items were found. The NRC Resident has been notified. | |
| ENS 56803 | 19 October 2023 16:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: On 10/19/2023, at approximately 1110 (CST), with Unit 1 in mode 1 at 100 percent power, the reactor automatically tripped. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The cause of the trip is being investigated. Operations responded and stabilized the plant. Auxiliary feedwater actuated as expected. Decay heat is being removed by the steam generator through the steam generator power operated relief valve. The trip was complex as non-safety related power was lost to both Unit 1 and Unit 2. Unit 1 is currently in mode 3 and on natural recirculation as both reactor coolant pumps are without power. Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was lost for approximately 70 minutes. No impacts to the SFP temperature were observed. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Due to the actuation of the auxiliary feedwater system following the reactor trip, this event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The second paragraph of the original report is amended as follows to correct information regarding the spent fuel pool for Unit 2: Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was maintained at all times with one train of SFP cooling. The second train lost power and was restarted approximately 70 minutes (after power was lost). No impacts to the SFP temperature were observed. Notified R3DO (Orth) and IR MOC (Crouch) and NRR EO (Felts) via email | |
| ENS 56804 | 19 October 2023 16:10:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Seismic Monitoring Capability | The following information was provided by the licensee via phone and email: Reporting due to loss of emergency preparedness capabilities. Seismic monitoring capability is non-functional due to loss of power. These monitors do not have a credited compensatory measure. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The NRC Resident Inspector has been notified. The licensee intends to notify state and local officials. | |
| ENS 56543 | 27 May 2023 23:34:00 | 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 | Notification of Unusual Event Due to Multiple Fire Alarms in Containment Not Verified within 15 Minutes | The following information was provided by the licensee via email: Notification of Unusual Event, HU4.1 declared based on multiple fire alarms in the containment building not verified within 15 minutes. Turbine trip causing reactor trip due to fault on 2GT transformer. At 1845 CDT, verification of no fire in the containment building. Notified DHS Senior Watch Officer, FEMA Operations Center, CISA Central watch officer, DOE Operations Center (email), HHS Operations Center (email), EPA Emergency Operations Center (email), USDA Operations Center (email), FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email), FEMA NRCC (email) and CWMD watch desk (email).
The following information was provided by the licensee via email: This update is being made to report the actuation of the auxiliary feedwater system following the reactor trip at 1819 CDT. This event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). This update is also being made for the termination of the notification of unusual event at 2304 CDT on 5/27/2023. The basis for the termination was that there was no indication of a fire. Upon lockout of 2GT transformer, main to reserve power transfer did not occur on 3 of 4 non-safeguards buses. Subsequently, operator action successfully restored power to all non-safeguards buses at 1925 CDT. There was no impact to the health and safety of the public or plant personnel. The NRC resident inspector has been notified of the update. Notified R3DO (Benjam¡n), NRR EO (Walker), IRMOC (Grant), DHS Senior Watch Officer, FEMA Operations Center, CISA Central watch officer, DOE Operations Center (email), HHS Operations Center (email), EPA Emergency Operations Center (email), USDA Operations Center (email), FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email), FEMA NRCC (email) and CWMD watch desk (email). | |
| ENS 56266 | 9 December 2022 04:01:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Agency Notification Due to Chemical Leak | The following information was provided by the licensee via email: On 12/8/2022, Prairie Island Nuclear Generating Plant initiated a notification to the State of Minnesota due to a HVAC coolant leak reaching waters of the state. The estimated quantity is 5 gallons of NALCO LCS-60. The leak was due to a failed heat exchanger coil and has been isolated. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: At 0019 EST on 12/9/2022, the Prairie Island Nuclear Generating Plant (PINGP) made Event Notification 56266 notifying the NRC of an environmental report to the State of Minnesota due to an estimated 5 gallons of NALCO LCS-60 that leaked from a failed heat exchanger coil and reached the waters of the state. This event notification was made in accordance with 10 CFR 50.72(b)(2)(xi). During further review of NRC reporting guidance, PINGP has concluded that the reported quantity of NALCO LCS-60 that leaked during this event was below the reporting threshold outlined in NUREG 1022, Revision 3. The NRC Resident Inspector has been notified. Notified R3DO (Kozak) | |
| ENS 55942 | 14 June 2022 13:47:00 | 10 CFR 26.719, FFD Reporting requirements | FITNESS-FOR-DUTY Report | The following information was provided by the licensee via email: A licensed operator supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant is on hold in accordance with the licensee's fitness-for-duty policy. The NRC Senior Resident Inspector has been notified. | |
| ENS 55529 | 18 October 2021 00:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Emergency Diesel Generators Simultaneously Inoperable | At 1930 CDT on 10/17/2021, it was discovered both of the Unit 2 Emergency Diesel Generators were simultaneously INOPERABLE with a requirement to have one OPERABLE train; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10CFR 50.72(b)(3)(v). Offsite power was OPERABLE during this event. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. The cause was corrected and both Emergency Diesel Generators are currently operable.
This is a retraction of Event Notification EN55529 in accordance with 10 CFR 50.72(b)(3)(v)(D) made by the Prairie Island Nuclear Generating Plant on October 18, 2021. The original notification stemmed from a loss of power to the non-safety related Unit 2 Emergency Diesel Generator (EDG) starting air compressors. The resulting pressure decay in the EDG starting air receivers led to a decision to declare both EDGs inoperable. A subsequent engineering evaluation has provided reasonable assurance that the Unit 2 EDGs were operable and capable of performing their safety function during the time power was lost. The NRC Resident Inspector has been notified. The HOO notified R3DO (Skokowski). | |
| ENS 55503 | 3 October 2021 20:25:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation | At 1525 CDT, 10/3/2021, with Unit 2 in Mode 5 at 0 percent power for a refueling outage, the 22 Turbine-Driven Auxiliary Feedwater (AFW) pump received an actuation signal during preparations for an Integrated Safety Injection test. The reason for the actuation signal is under investigation. The AFW steam admission valve opened and then, due to plant conditions, received a trip signal due to low discharge pressure. The steam supplies to the TD AFW pump were isolated. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. Unit 1 was not affected by this issue. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 55377 | 23 July 2021 15:40:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Fish Kill | At approximately 1040 CDT, July 23, 2021, the Minnesota State Duty Officer was notified by Xcel Energy Environmental Services of a fish kill in the Prairie Island Nuclear Generating Plant discharge canal. The fish kill resulted from a change in temperature due to the loss of power to the plant cooling tower pumps. The cause of the power loss is under investigation. This notification is being made as a four-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 54877 | 3 September 2020 19:21:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to a Hydraulic Oil Spill | On 9/3/2020, at 1421 (CDT), a notification was made to the National Response Center as a result of a spill of approximately 3 fluid ounces of hydraulic oil into the plant waters of Prairie Island Nuclear Plant intake on 9/2/2020. This notification is being made as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact to public health and safety. The NRC resident inspector has been notified. | |
| ENS 54859 | 26 August 2020 18:19:00 | 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 Flux Rate | At 1319 CDT, on August 26, 2020, with Unit 1 in Mode 1 at 95.2 percent power in coast down for the 1R32 refueling outage, the reactor automatically tripped due to flux rate. All systems responded normally to these conditions with auxiliary feedwater initiating as expected. Operations stabilized the plant without complication. Decay heat is being removed via a main feedwater pump to the steam generators. Unit 2 is not affected and remains at 100 percent power. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour report per 10 CFR 50.72(b)(3)(iv)(A), specified system actuation. The cause of the scram is under investigation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 54803 | 29 July 2020 00:49:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Failed Radiation Detector | Radiation monitor RE-69, 'Guardhouse Area Radiation Detector' was found failed on operator rounds. RE-69 is a category A1 piece of equipment that provides the sole means of indication for a parameter used to directly assess emergency action level (EAL) RA3.1, 'dose rate greater than 15 mR/hr in either of the following areas: Control Room (R-1), Central Alarm Station (R-69).' A compensatory measure was subsequently established and the ability to assess EAL RA3.1 has been restored. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. | |
| ENS 53853 | 31 January 2019 13:43:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | En Revision Imported Date 3/25/2019 | EN Revision Text: BOTH EMERGENCY DIESEL GENERATORS INOPERABLE DUE TO LOW AIR TEMPERATURE At 0743 (CST) on 1/31/2019, both trains of Unit 2 Diesel Generators were declared INOPERABLE due to outside air temperature exceeding the low temperature design limit for the diesel engines; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v) for an event or condition that could have prevented the fulfillment of a safety function. The Unit 2 Diesel Generators are still able to start if necessary to provide power. Additionally, multiple layers of defense in depth measures are in place to ensure safety. Prairie Island has five sources of offsite power; all of which are currently available. The Unit 1 Diesel Generators are OPERABLE and capable of being cross-connected to Unit 2. Additional equipment capable of responding to beyond design basis events is available on site providing another layer of defense in depth. Both Unit 2 Diesel Generators were returned to an OPERABLE status at 0810 on 1/31/2019 based on outside air temperature rising above the low temperature design limit with forecasted temperatures to remain above the low temperature design limit. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The air temperature limit was -30 degrees Fahrenheit. Unit 1 was not affected. The EDGs were supplied by a different manufacturer with different air temperature limits.
Engineering analysis performed subsequent to the event notification has determined that both Unit 2 Diesel Generators would have been able to fulfill their safety function during the period of time when the outside air temperature had exceeded the low temperature design limit. Therefore, EN# 53853 is being retracted. The NRC Resident Inspector has been notified of the event notification retraction. Notified R3DO (McCraw). | |
| ENS 53757 | 28 November 2018 06:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Inadvertent Siren Activation | At 0752 CST, on November 28, 2018, Dakota County inadvertently actuated their sirens while performing a scheduled weekly (Emergency Planning Fixed Siren Test). All seven (7) Dakota County sirens actuated for approximately 9 seconds before Dakota County Dispatch canceled the activation. This 4-hour non-emergency report is being made per 10 CFR 50.72(b)(2)(xi), Offsite Notification (which was made to Dakota County Dispatch). Capability to notify the public was never degraded during this time. All Emergency Notification sirens remain in service. No press release is planned at this time. The licensee has notified the NRC Resident Inspector. | |
| ENS 53700 | 28 October 2018 05:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Seismic Monitoring Panel System Inoperable | This event is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) for a major loss of emergency assessment capability at the Prairie Island Nuclear Generating Plant. At 1435 CDT on October 28, 2018, troubleshooting of the Seismic Monitoring Panel resulting from the receipt of Control Room annunciator 47023-0603 (Seismic Monitor Panel) determined that the '(Operational Basis Earthquake) OBE Exceedance' alarm on the Seismic Monitoring Panel will not alarm and determined the panel is non-functional. The Seismic Monitoring Panel system functions to provide indication that the OBE threshold has been exceeded following a seismic event and is used in the Prairie Island Nuclear Generating Plant Emergency Plan to perform classification of Initiating Condition 'Seismic event greater than OBE levels' and Emergency Action Level HU2.1. Station personnel are monitoring the seismic recorders for event alarms on a 15 minute frequency due to alarm function failure. The station is developing repair plans for restoration of the alarm function. This event does not adversely affect the safe operation of the plant or health and safety of the public. The licensee has notified the NRC Resident Inspector. | |
| ENS 53546 | 10 August 2018 05:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Emergency Diesel Generator Cooling Water Pumps Declared Inoperable | On 8/10/2018 at 1445 (CDT) both trains of Cooling Water (Cooling Water Pumps for Emergency Diesel Generators) were declared INOPERABLE and both units entered (Technical Specification) (TS) 3.0.3 due to corroded jacket cooling water plugs for (the) 12 and 22 cooling water pump motors; therefore this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). At 1543 (CDT), 08/10/2018 the 121 Cooling Water pump was aligned to the "A" Cooling Water train and the TS 3.0.3 condition was exited for both units. (After restoring train A cooling water the site entered a seven day limiting condition for operations, TS 3.7.8 for one inoperable cooling water pump.) There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
Testing and forensic analysis performed subsequent to the notification has determined the as-found condition would not have impacted either diesel-driven pumps' ability to start, run, and meet flow/pressure requirements to perform their required safety function. Therefore, EN# 53546 is being retracted. The NRC Resident Inspector has been notified of the Event Notification retraction. Notified R3DO (Kozak). | |
| ENS 53408 | 17 May 2018 16:15:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Emergency Diesel Generator | At 1115 (CDT) on May 17, 2018 with Unit 2 in Mode 1 at 100% power, the station experienced an auto-start of Emergency Diesel Generator, D5. Preliminary information indicates that the Bus 25 Potential Transformer (PT) fuse drawer was inadvertently opened, causing Breaker 25-16 to open and de-energize bus 25. Operators were able to manually close the D5 EDG output breaker to re-energize bus 25. 22 Component Cooling (CC) Pump auto-started on the loss of 21 CC Pump due to low pressure in the CC system as designed. All equipment functioned as designed. 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 EDG. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. | |
| ENS 53286 | 23 March 2018 21:35:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Assessment Capability | This report is made for a loss of Emergency Assessment Capability associated with Emergency Action Levels for Toxic and Flammable Gas and is reportable under 10 CFR 50.72 (b)(3)(xiii). During an emergency equipment inventory it was identified that methods were not available to detect levels of toxic or flammable gas at the IDLH (Immediately Dangerous to Life and Health) level for a number of substances. The IDLH is used to assess the Alert Emergency Action Level. The ability of the Control Room Staff to detect and respond to the presence of toxic or flammable gas is unaffected. Because there have been no chemical spills or releases that would require sampling to be performed, the health and safety of the public was not affected. The NRC Resident Inspector has been notified. The State of Minnesota will be notified. | |
| ENS 53067 | 12 November 2017 09:03:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Containment Spray Pump Control Switches Out of Service | At 2119 (CST) on 11/12/2017 a Control Room board walk down discovered that both of the Unit 2 Containment Spray Pump control switches were in pull-out. With the control switches in pull-out, the pumps would not automatically start as required. Unplanned TS (Technical Specifications) 3.0.3 was entered at 2119 as a result of not complying with TS 3.6.5, Containment Spray and Cooling Systems, which requires both trains of Containment Spray to be Operable while in Mode 4. Unit 2 had entered Mode 4 at 0303 on 11/12/2017. TS 3.0.3 was exited at 2127 on 11/12/2017 when both Containment Spray Pump control switches were placed in Automatic restoring Operability. Preliminary investigation determined that while Unit 2 was in Mode 5, Surveillance SP 2099, Main Steam Isolation Valve Logic Test, had taken the Containment Spray Pump control switches to pull-out but did not re-align the control switches to automatic after the test was complete. This 8-hour Non-Emergency report is being made per 10 CFR 50.72(b)(3)(v)(D), Accident Mitigation. The NRC Senior Resident Inspector has been informed. | |
| ENS 53135 | 26 October 2017 01:50:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Cooling Water System | At 2050 (CDT) on October 25, 2017, during the Unit 2 refueling outage, with no fuel in the Reactor Vessel, control room operators found that Unit 1 Train B Containment Fan Coil Units (FCUs) had swapped from chilled water to cooling water (CL). Construction Electricians were installing a new relay 2Sl-22X when the plunger on adjacent relay 2Sl-23X was bumped, which caused the swap of the Unit 1 Containment Fan Coil Units (CFCUs) from chilled water to cooling water. Relay 2SI-23X is a slave relay that starts 22 Turbine Driven Auxiliary Feed Water Pump, illuminates blue lights on various control switches, closes MV-32159 Loop A/B CLG WTR HDR XOVR MV B, closes chilled water Isolation Valves to Unit 1 Train B, and closes chilled water Isolation Valves to Unit 2 Train B. This actuation was as expected. CL is a shared system and, upon a Safety Injection (SI) signal on either unit, the CL header splits into two trains and, as a result, the CL supply is isolated to the chillers that supply chilled water to both units' CFCUs. By design, CL is the safety related source of cooling to the CFCUs. 22 Turbine Driven Auxiliary Feed Water Pump did not start as the unit was in 'No Mode' with the control switch for the pump in manual. MV-32159 did automatically close per design. Unit 2 Chilled Water was already isolated due to work in progress with the unit in 'No Mode.' There was no impact on public health and safety. This event is being reported as a 60-day telephone notification in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an invalid actuation of a containment heat removal system (the FCUs were running, but were swapped to their safeguards source due to an invalid actuation of a relay). The licensee notified the NRC Resident Inspector. | |
| ENS 53134 | 23 October 2017 13:56:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Emergency Diesel Generator | At 0856 (CDT) on October 23, 2017, during the Unit 2 refueling outage, with no fuel in the Reactor Vessel, an unexpected auto start of the Unit 2 Train B Emergency Diesel Generator (D6) occurred when construction electricians inadvertently bumped the plunger for relay 2Sl-20X while working in the relay rack. Relay 2Sl-20X is a slave relay that actuates a light on the control board, starts D6, and starts 22 Residual Heat Removal (RHR) pump on a Safety Injection signal. In this instance, the RHR pump did not start as its control switch was in pull-out. It is expected that the control board light lit for the brief time the relay plunger was depressed, but this could not be confirmed. The D6 actuation resulted in an unexpected annunciator for D6 EMERGENCY GENERATOR SI SIGNAL EMERGENCY START. Operators responded per the alarm response procedure, performed a walk down of running D6 and then performed a shutdown of D6. D6 started and functioned as expected. There was no impact on public health and safety. This event is being reported as a 60-day telephone notification in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an invalid actuation of an emergency diesel generator. The NRC Resident Inspector has been notified. | |
| ENS 53019 | 16 October 2017 19:25:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System Shutdown Communication Line Vent Through Wall Defect | At 1425 CDT on 10/16/17, investigation into a boric acid indication was determined to be a through wall leak at the socket weld that joins the 3/4 inch line 2RC-92 to valve 2RC-8-37. Unit 2 is currently in Mode 5 with Reactor Coolant system (RCS) Operational Leakage limits not applicable. The leak is downstream of two first off RCS isolation valves that are normally closed. The leak is not quantifiable as it only consists of a small amount of dry boric acid at the location. This failure constitutes welding or material defects in the primary coolant system that cannot be found acceptable under ASME Section Xl. Therefore, this is a degraded condition reportable under 10 CFR 50.72(b)(3)(ii)(A). At the time of this notification, the Prairie Island Nuclear Generating Plant Unit 2 is in Mode 5 for a planned refueling outage. The identified defect will be repaired prior to entering Mode 4. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. | |
| ENS 52742 | 10 May 2017 12:55:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Inadvertent Siren Actuation | At approximately 0755 CDT, on May 10, 2017, Pierce County inadvertently actuated their sirens while performing a scheduled weekly cancel test. All fifty two (52) Pierce County sirens actuated county wide for approximately 11 seconds before Pierce County Dispatch canceled the activation. This 4-hour non-emergency report is being made per 10 CFR 50.72(b)(2)(xi), Offsite Notification. Capability to notify the public was never degraded during this time. All Emergency Notification sirens remain in service. No press release is planned at this time. The license has notified the NRC Senior Resident Inspector. | |
| ENS 52638 | 20 March 2017 21:39:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Historical Event Related to Past Operability | On February 3, 2017, Prairie Island staff performed maintenance on the transom above Battery Room Door 225. This activity resulted in the transom being unlatched for approximately five minutes. On February 6, 2017, a question from the NRC Resident Inspector resulted in an evaluation of this condition for past operability. On March 20, 2017, the past operability evaluation of Door 225 concluded that, in the event of a postulated HELB (High Energy Line Break), the transom being unlatched during the five minute maintenance period resulted in the inoperability of multiple systems in the Unit 1 and Unit 2 battery, auxiliary feedwater, and Unit 1 safeguards bus rooms that would be required to mitigate the postulated HELB. The loss of safety functions required to mitigate the postulated HELB make the condition reportable under 50.72(b)(3)(ii) for an unanalyzed condition that significantly degrades plant safety. Unlatching the transom above the Battery Room Door creates an opening not accounted for in design bases documents. This occurred due to an improperly prepared work permit. Corrective actions are in place to preclude recurrence. The licensee informed the NRC Resident Inspector.
Further analysis determined that an unlatched transom would result in a relative humidity of 100 percent in 11 Battery Room for about 10 minutes following a postulated HELB. Since the equipment in the Battery Rooms is not qualified for a harsh environment, the components in 11 Battery Room would have been inoperable. Temperature and relative humidity in the other Battery Rooms, Auxiliary Feedwater Rooms, and the Unit 1 Safeguards Bus Rooms would have remained within the allowable limits. Therefore, for the five minutes the strike was removed from the transom, only equipment in 11 Battery Room and supported A Train components would have been inoperable. This event was not an Unanalyzed Condition that significantly degraded plant safety, under 10 CFR 50.72(b)(3)(ii), as no safety function would have been lost. The licensee notified the NRC Resident Inspector. Notified R3DO (Skokowski). | |
| ENS 52474 | 4 January 2017 19:47:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty | A non-licensed employee supervisor had a confirmed positive for a prohibited substance during a random fitness-for-duty test. The individual's unescorted access to the plant has been denied. The licensee notified the NRC Resident Inspector. | |
| ENS 52178 | 14 August 2016 04:59:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Excessive Reactor Coolant System Leakage | Prairie Island Unit 1 declared an Unusual Event at 2359 CDT on 8/13/2016 based on Reactor Coolant System (RCS) identified leakage being greater than 25 gpm. The RCS leakage was 40 gpm for three (3) minutes. The RCS leakage was stopped when letdown flow was isolated. Minimum charging flow has been established and Excess Letdown was placed in service. Prairie Island Unit 1 is currently stable and continues to operate at 100 percent power. There was no impact on Prairie Island Unit 2. CV-31339 (Letdown Line Containment Isolation Valve) failed closed. VC-26-1 (Regenerative Heat Exchanger Letdown Line Outlet Relief to Pressurizer Relief Tank (PRT)) lifted with 40 gallons per minute to the PRT for three (3) Minutes. Operators entered procedure 1C12.1 AOP3, Loss of Letdown Flow to VCT. Letdown was isolated per 1C12.1 AOP3, relief valve VC-26-1 reseated and leakage to the PRT stopped. Charging flow was reduced to one (1) charging pump at minimum speed (16 GPM). Excess letdown was placed in service to maintain pressurizer level between 32 - 34 percent. The cause for CV-31339 closing has not yet been determined. The NRC Resident Inspector has been notified. Notified DHS SWO, FEMA Ops Center, DHS NICC. Notified FEMA National Watch and Nuclear SSA via E-mail.
At 0329 CDT the Notice of Unusual Event was terminated based on confirmation that conditions meet all termination criteria. RCS conditions are stable. RCS leakage is less than Technical Specification limits. The current value (of RCS identified leakage) is 0.038 gpm. No classification criteria is currently met. The NRC Resident Inspector has been notified. Notified R3DO (Kozak), NRR EO (Miller), and IRD (Stapleton). Notified DHS SWO, FEMA Ops Center, DHS NICC. Notified FEMA National Watch and Nuclear SSA via E-mail. | |
| ENS 51877 | 21 April 2016 20:50:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Missing Fire Barriers | Missing fire barrier between Fire Area (FA) 59 and 85. During a walk down of fire barriers for the NFPA 805 project, it was determined that the fire barrier between Fire Area 59 (Unit 1) and 85 (common) is not a rated barrier due to unsealed penetrations in the barrier. Evaluation FPEE 12-006 evaluated the acceptability of the barrier being unrated based on separation of safe shutdown equipment however a review of equipment credited for Appendix R safe shutdown identified that the redundant credited Appendix R equipment is on either side of the fire barrier which is not 3 hour rated. The conclusion of the FPEE is therefore no longer valid. Fire Hazard Analysis Drawings Do Not Match Boundary Description. The plant layout in F5 Appendix F, Rev. 28, Fire Hazard Analysis (FHA), does not agree with the boundary description in the FHA for the Unit 1 and 2 Containment Annulus fire areas, Fire Area (FA) 68 and 72. The layout should but does not show the fire area boundary between the annulus and adjacent fire areas, FA 60 and 75 on 735 (foot) and 61A on 755 (foot), as an Appendix R boundary. The annulus airlock doors are 3-hour fire rated and the airlock is constructed of concrete thick enough to qualify as a 3 hour fire barrier however, there are penetrations in the barrier that are not sealed with fire rated materials or inspected as required by the Fire Protection Program. Therefore, this is an unanalyzed condition reportable under 10 CFR 50.72(b)(3)(ii)(B). This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. | 05000282/LER-2016-004 |
| ENS 51840 | 31 March 2016 08:24:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Prohibitied Item in the Protected Area | On 3/31/2016 at approximately 0342 CDT, a worker within the Protected Area self-reported a can of beer had been packed in the worker's lunchbox. The worker reported after opening the can and taking a sip it was discovered to be a beer. This event is reportable under 10 CFR 26.719(b)(1). The worker notified Security who immediately escorted the worker from the Protected Area and disposed of the beer. The worker is not an Operator or a Supervisor. The investigation of this event is in progress. The public health and safety are not impacted. The NRC Resident Inspector was notified. | |
| ENS 51642 | 8 January 2016 01:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition - Appendix R Credited Procedures Not in Place | Prairie Island's Appendix R calculations credit a procedurally established repair instruction to the Train B Pressurizer Vent valves for a postulated fire in Fire Area 59 (Unit 1) and Fire Area 74 (Unit 2) to obtain Mode 5 during a postulated fire in the affected areas. At 1900 (CST) on 1/7/2016, during a review of corrective actions associated with Prairie Island's NFPA 805 transition, it was identified that the required procedures are not in place to make the analyzed repairs. It has been determined that this condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. As compensatory measures, hourly fire watches are in place in the affected areas of the Auxiliary Building. The operating crew and Fire Brigade have been briefed on the impact of a fire in the affected area. This brief will continue to future operating shifts via a standing instruction. Fire detection equipment for the affected zones has been protected to ensure availability and operating crews are walking down the affected areas to verify any required transient combustibles in the affected areas are controlled in accordance with plant procedure. These compensatory measures, in addition to automatic fire detection and suppression capability in these fire areas, ensure protection of the potentially affected equipment until corrective actions can be completed. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. | 05000282/LER-2016-003 |
| ENS 51616 | 21 December 2015 22:25:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Non-Compliant Fire Protection Manual Operator Actions | As part of the License Amendment development to transition to NFPA 805, PINGP (Prairie Island Nuclear Generating Plant) Calculation ENG-ME-353, Mechanical MOV (Motor Operated Valve) Analysis to support IN-92-18 Response, revision 1, issued in 1998, was reviewed for applicability for the transition to NFPA 805. Recent consultation with an MOV engineer regarding the scope of the revision indicated ENG-ME-353 is out of date. On 12/21/2015, during technical review for a new weak link calculation, several MOVs were identified from the list of MOVs that are credited to be manually operated from outside the control room in the event of a fire in the control room or relay room per PINGP Procedure F5 Appendix B, Control Room Evacuation (Fire), that could be damaged if hot shorts were to bypass the torque and limit switches. There are also four other motor valves associated with the Gland Steam system of both Unit 1 and Unit 2 that were added to the procedure F5 Appendix B, Control Room Evacuation (Fire), that have not been analyzed for a weak link. This unanalyzed condition could impact the ability of plant operators to implement procedure F5 Appendix B, Control Room Evacuation (Fire). New hourly fire watch impairments were created for Fire Area 13 (Control Room) and Fire Area 18 (Relay and Cable Spreading Room) as compensatory measures. Therefore, this is an unanalyzed condition reportable under 10 CFR 50.72(b)(3)(ii)(B). The public health and safety is not impacted. The NRC Resident Inspector has been notified.
Reviews of the list of MOVs susceptible to hot shorts bypassing the torque and limit switches credited to be manually operated from outside the control room in the event of a fire have continued. Additional valves have been noted to be affected by this failure mechanism in areas outside of the Control Room or Relay Room. The additional MOVs affected by this unanalyzed condition could impact the ability of plant operators to implement PINGP Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room. As a compensatory measure, additional hourly fire watch impairments were created for the following fire areas: Fire Area 031 ( A Train Hot Shutdown Panel & Air Compressor/Aux 695 Feedwater Room) Fire Area 032 ( B Train Hot Shutdown Panel & Air Compressor/Aux 695 Feedwater Room) Fire Area 058 (Aux Building Ground Floor Unit 1) Fire Area 073 (Auxiliary Building Ground Floor Unit 2) The public health and safety is not impacted. The (NRC) Resident Inspector has been notified. Notified R3DO (Duncan). | 05000282/LER-2016-001 |
| ENS 51609 | 17 December 2015 19:18:00 | 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 | Unusual Event Declared Due to Fire Alarm in Containment Not Verified within 15 Minutes | Unusual Event HU2.1 declared at 1318 (CST). A fire alarm was received in unit 2 containment at 1307 (CST). Due to the location of the alarm, personnel were unable to verify the status within 15 minutes. At 1343 (CST), the fire alarm in containment cleared. This alarm came in shortly after a unit 2 reactor trip. The reactor trip was due to a turbine trip. Decay heat removal is via forced circulation with aux feed and steam dumps providing secondary cooling. Offsite power remains available. The reactor trip was uncomplicated and all control rods inserted. 25B feedwater heater relief valve lifted and has reseated. No offsite assistance was requested. The licensee has notified the NRC Resident Inspector. State and local authorities were notified.
The licensee terminated the NOUE (Notification of Unusual Event) at 1450 CST. The basis for the termination was determination that there was no smoke or fire in the Unit 2 containment observed during containment entry. NRC Resident Inspectors were notified. State and local governments were notified. The health and safety of the public was not at risk. Notified the R3DO (Valos), NRR EO (Morris), IRD (Grant), DHS SWO, FEMA Ops enter, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA. | |
| ENS 51528 | 11 November 2015 14:26:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Shield Building Vent Gas Radiation Monitor Out of Service for Planned Maintenance | At 0826 CST on 11/11/2015, 1R-22, Shield Building Vent Gas Radiation Monitor, was removed from service for planned maintenance. This monitor has no compensatory measure that will allow timely classification of two Emergency Action Levels (EALs) - NUE (Notification of Unusual Event) and Alert classifications - when out of service. It is also used for offsite dose projection calculations. This results in a Loss of Emergency Assessment Capability while 1R-22 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii). Unit 1 Shield Building Ventilation Stack is also monitored by high range monitor, 1R-50, which is used for the same purpose in Site Area or General Emergency classifications. 1R-50 is being monitored and is indicating normal values. There are no radioactive leaks that will impact the Shield Building as evidenced by normal readings on 1R-22 prior to its removal from service. The duration of this maintenance is scheduled for 24 hours and will continue until the monitor is returned to service. Maintenance will not result in the unplanned release of radioactivity to the environment and will not adversely affect the safe operation of the plant or health and safety of the public. The licensee has notified the NRC Resident Inspector.
The licensee indicated that the duration of maintenance was extended for approximately 24 hours to allow continued repair of the monitor. The NRC Resident Inspector was notified. Notified the R3DO (Kozak). | |
| ENS 51498 | 26 October 2015 12:03:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Associated with a High Energy Line Break | At 0703 CDT on 10/26/2015, Prairie Island Nuclear Generating Plant (PINGP) identified Door 62, '11/21 Auxiliary Feedwater Pump (AFW) Room to 12/22 AFW Pump Room' to be in a closed position. Door 62 functions as a fire door, and closes in the event of a fire in either A or B Train AFW Pump Rooms. When Unit 1 or Unit 2 is in modes 1-4, Door 62 is required to remain open in the event of an internal flood in AFW Pump Room or a Turbine Building High Energy Line Break (HELB) Flood into the AFW Pump. Currently, Unit 2 is in MODE 6, so only the Unit 1 Turbine Building HELB applies. Door 62 is normally open with a fusible link to allow closure during a fire. Door 62 was closed during maintenance. With the door closed the Unit 1 side of the Auxiliary Feedwater Pump Room lost its ability to adequately drain water in a Unit 1 HELB event and was in an unanalyzed condition. Upon discovery, Door 62 was immediately repositioned to be open per the analyzed condition and a fire watch established per plant procedure. This notification is being conducted in accordance with 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition. The plant remains safe, and this condition does not pose any additional risk to the public. The licensee has notified the NRC Resident Inspector.
Prairie Island Nuclear Generating Plant is retracting this event notification, EN# 51498. Further analysis determined that the closure of Door 62 would not have prevented the structures, systems and components (SSC) located in the AFW Pump rooms, or SSCs powered from Motor Control Centers (MCCs) located in the AFW Pump rooms, from performing their safety functions. This is because the door closure would not have caused water level to rise above the maximum tolerable water height during any design basis flooding event. The acceptance criteria of the area flooding calculations were still met with Door 62 closed. Therefore, the Unit 1 side of the AFW Pump room did not lose its ability to adequately drain water in a Unit 1 HELB event, this event was not an 8-hour notification for an Unanalyzed Condition that significantly degrades plant safety, under 10CFR50.72(b)(3)(ii). The licensee has notified the NRC Resident Inspector. Notified R3DO (Lipa). | |
| ENS 51497 | 24 October 2015 11:57:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification for Onsite Incident of Local Interest | Prairie Island Nuclear Generating Plant (PINGP) notified Local Law Enforcement Agency and the Prairie Island Community Tribal Council due to an incident onsite of local interest. This notification is being conducted in accordance with 10 CFR 50.72(b)(2)(xi) for notification to an outside government agency. The plant remains safe, and this condition does not pose any additional risk to the public. The licensee has notified the NRC Resident Inspector. | |
| ENS 51429 | 28 September 2015 18:27:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Unit 1 Emergency Diesel Generators Inoperable at the Same Time | At approximately 1327 CDT on September 28, 2015, both D1 and D2 Diesel Generators (EDG) were inoperable simultaneously until corrected at 1345 CDT. The D2 Diesel Generator had been declared inoperable for the planned performance of SP1307, D2 Diesel Generator 6 Month Fast Start Test. Tech Spec LCO 3.8.1 Condition B had been entered for D2 Diesel Generator. Subsequently, D1 Diesel Generator was determined to be inoperable but available due to Train A Cooling Water Header being inoperable during post maintenance testing of SV-33133, Backwash Water Supply to the 121 Safeguards Traveling Screen. Tech Spec LCO 3.7.8 Condition B was entered for the Cooling Water Header inoperability, which forced a cascade to Tech Spec 3.8.1 Condition B for D1 Diesel Generator. With both Emergency Diesel Generators inoperable, Tech Spec 3.8.1 Condition E was entered, which required the restoration of one Emergency Diesel Generator to operable status within 2 hours. D2 was returned to operable status through completion of SP 1307, and Tech Spec 3.8.1 Condition E was exited at 1345 CDT. With both Emergency Diesel Generators inoperable, this condition is reportable under 10 CFR 50.72 (b)(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. The plant remains safe, and this condition does not pose any additional risk to the public. Additionally, our defense in depth strategies are relied upon to take actions to protect the health and safety of the public. D2 Diesel Generator remained available with full cooling water flow during this time. The safety significance of this event is low, as engineering hydraulic analysis has demonstrated that with the safeguards traveling screen backwash water supply valve fully opened, the Cooling Water System would have continued to provide full cooling flow to the D1 Diesel Generator. The licensee has notified the NRC Resident Inspector.
An evaluation has been performed and it has been determined that SV-33133 and SV-33134 do not have an active close safety function. The Cooling Water System analysis of record, calculation ENG-ME-820, Rev 0B shows that the Cooling Water System continues to have flow margin with screen wash control valves SV-33133 and SV-33134 open. Therefore, there is no need for the valves to close to ensure the Cooling Water System's safety function. Because the valves do not have a safety function to close, this event was not an event or condition which could have prevented the fulfillment of a safety function of an SSC (structures, systems and components) required to mitigate the consequences of an accident and, therefore, did not require an 8 hour notification in accordance with 10 CFR 50.72(b)(3)(v)(D) for an event or condition that could have prevented fulfillment of a safety function (i.e., accident mitigation) under 10 CFR 50.72(b)(3)(v)(D). The notification is hereby retracted. The licensee has notified the NRC Resident Inspector." Notified R3DO (Orlikowski). | |
| ENS 51266 | 27 July 2015 14:02:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Outage for Plant Computer Systems | On July 27, 2015 at 0902 (CDT), (the site commenced) a planned outage of the Emergency Response Data System (ERDS) and Safety Parameter Display System (SPDS), referred to as 'plant computer'. The unavailability of ERDS and SPDS could significantly affect the site's ability to respond to an emergency if one were to occur. During this time, Operations will be utilizing the site's procedures 1C1.5 and 2C1.5, 'OPERATION WITHOUT COMPUTER', which requires additional operators for monitoring of equipment affected by the loss of the plant computer. Additionally, as this is a planned outage, the work week schedule has been modified to ensure limited interactions required by Operations during this time frame. The site expects ERDS and SPDS to be operational 1200 July 28, 2015. This event is reportable under 10 CFR 50.72(b)(3)(xiii), 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 (ENS), or Offsite Notification System). The ENS and Offsite Notification System are not affected by this planned outage. The health and safety of the public are not impacted by this planned outage. The NRC Resident Inspector has been informed. | |
| ENS 51136 | 7 June 2015 12:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Resulting from Turbine Trip on Low Oil Pressure | The following was received via phone and fax: On June 7, 2015, at 0735 CDT, the Unit 2 Reactor automatically tripped while operating at 100 percent power due to an automatic Turbine trip due to low bearing oil pressure. The crew entered the reactor trip emergency operating procedure and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The automatic Reactor trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the Auxiliary Feedwater Pumps as designed on low narrow range Steam Generator level and provided makeup flow to the Steam Generators. The Auxiliary Feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam Generator levels have been returned to normal. Auxiliary Feedwater has been secured. Steam Generators are being supplied by (the) 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of the Turbine trip remains under investigation. There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified. | |
| ENS 51107 | 1 June 2015 03:20:00 | 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 Trip of Condensate and Main Feedwater Pump | On May 31, 2015 at 2220 CDT, the Unit 1 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 11 Condensate Pump followed by a lockout trip of 11 Main Feedwater Pump. Manual Reactor Trip is directed by the annunciator response procedure for the lockout alarm, C47010-0101, 11 Feedwater Pump Locked Out. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. Following the reactor trip, 15A Feedwater Heater relief lifted and failed to reseat. 12 Main Feedwater Pump was subsequently secured resulting in 15A Feedwater Heater relief valve reseating successfully. Steam generators are being supplied by 12 Motor Drive Auxiliary Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of 11 Condensate Pump trip remains under investigation. There was no effect on Unit 2 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified. Unit 2 is unaffected and remains at 100 percent power. | |
| ENS 51013 | 25 April 2015 02:25:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Assessment Capability Due to Out-Of-Service Radiation Monitors | At 2125 on 4/24/15, 1R-2, Containment Vessel Area Radiation Detector failed. Previously, 1R-7, Incore Seal Table Area Radiation Detector, had failed on 4/20/15. The compensatory measure for 1R-2 out-of-service is to verify 1R-7 operating properly and the compensatory measure for 1R-7 out-of-service is to verify 1R-2 operating properly. With both monitors out-of-service and Unit 1 operating in Mode 5, no compensatory measure is available that will allow timely classification of two Emergency Action Levels (EALs) - Notification of Unusual Event (NUE) classification (RU2.2) and Alert classification (RA3.2). This results in a Loss of Emergency Assessment Capability while 1R-2 and 1R-7 are concurrently out-of-service. This is a reportable condition per 10 CFR 50.72(b)(3)(xiii). Monitoring of radiological conditions in Unit 1 Containment showed no indication of RCS leakage or elevated radiation levels prior to the failure of 1R-2. Unit 1 Containment also remains monitored by 1R-48, Containment Hi Range Area Radiation Detector A and 1R-49, Containment Hi Range Area Radiation Detector B, which currently indicate normal radiation levels. Unit 1 Shield Building Stack is also monitored by 1R-50, Shield Building High Range Vent Gas Radiation Detector, which also currently indicates normal radiation levels. Additionally, a temporary portable radiation monitor has been placed near the location of 1R-2 and is being continuously monitored. The plant remains in a safe condition and there was no effect to the health and safety of the public. The licensee has notified the NRC Resident Inspector. | |
| ENS 50994 | 19 April 2015 22:45:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Non Compliance with 10 Cfr 50 App R | On April 19, 2015, it was determined that in the event of an Appendix-R fire in the Control Room/Relay Room, fire induced circuit damage can potentially result in Reactor Coolant Pump's (RCP) restarting. Procedure F5 APP B does not take actions to open RCP DC knife switches as required per Appendix R calculation GEN-PI-026, GEN-PI-054, and GEN-PI-055. Neither unit is currently susceptible to this condition due to the installation of improved RCP seals which allow adequate time to restore seal cooling prior to seal failure. However, the condition has occurred within three years of the time of discovery and is reportable under 10 CFR 50.72(b)(3)(ii)(B). The Appendix R analysis requires that manual actions are taken to ensure that the RCPs are tripped and actions are taken to prevent restarting of the RCPs. The RCPs breaker must be verified open at the associated bus and DC knife switches located in the breaker cubicles are to be opened. These actions are required due to fire induced, loss of remote trip (spurious breaker closure) and loss of RCP seal cooling water and loss of Component Cooling (CC) water to the thermal barrier heat exchanger. These actions are required to achieve and maintain Mode 3, Hot Standby in the event of a catastrophic fire that results in the functional loss and/or evacuation of the Control Room/Relay Room. Current procedures verify the RCP breakers are open; however, they do not open the DC knife switch and therefore a hot short could result in a RCP restarting. This procedure omission meets the reporting criteria for 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Senior Resident Inspector. | |
| ENS 50966 | 7 April 2015 22:45:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Inadvertent Emergency Siren Activation | At approximately 1833 CDT on April 07, 2015, the licensee was notified by the siren vendor (NELCOM) that Dakota County Dispatch reported an inadvertent activation of an emergency siren (D-1), in Dakota County, MN. The cause of the siren activation is unknown. The siren was deactivated at 1810 CDT after sounding for approximately 25 minutes. The siren vendor (NELCOM) has been contacted to repair the siren. The siren remains out of service. This is the only siren out of service within the 10 mile Emergency Planning Zone (EPZ). There are 123 Emergency Notification sirens. Of the seven (7) credited sirens in Dakota County, six (6) remain in service. Thus, emergency notification capabilities remain in effect. NRC Resident has been informed. | |
| ENS 50950 | 3 April 2015 11:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Loss of Main Feedwater Pump | On April 3, 2015 at 0652 CDT, the Unit 2 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 21 Main Feedwater Pump as required by the annunciator response procedure for the lockout alarm. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. The auxiliary feedwater pumps have subsequently been secured and returned to automatic. Steam generators are being supplied by 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of 21 Main Feedwater Pump trip has been determined to be a failed suction pressure switch. There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified. The licensee plans to issue a press release. | |
| ENS 50948 | 2 April 2015 17:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Seismic Monitor Not Available for Emergency Plan Assessment | Northern States Power Company - Minnesota (NSPM) has completed a review of seismic monitor performance at the Prairie Island Nuclear Generating Plant (PINGP) over the past 3 years. The emergency preparedness plan requires seismic monitoring instruments to diagnose an earthquake for emergency action levels (EAL) HA1.1 (Seismic Event Greater Than Operating Basis Earthquake (OBE) as indicated by 'OBE Exceedance' alarm on Seismic Monitoring Panel) or HU1.1 (Earthquake felt in plant as indicated by Valid 'Event' alarm on Seismic Monitoring Panel). Contrary to that requirement, this review identified 6 unplanned instances where the seismic monitor was non-functional that were not previously reported, and 3 planned instances where the seismic monitor was non-functional for greater than 24 hours that were not previously reported. Since there was no compensatory measure that could be credited when the seismic monitor was non-functional, an emergency classification at the ALERT or UNUSUAL EVENT level could not be obtained with site instrumentation for a seismic event. The seismic monitor is currently functional, however it was determined to be non-functional on the following dates: Unplanned out of service: 1. August 14, 2012 2. November 16, 2012 3. November 18 2012 4. November 21, 2012 5. December 5, 2012 6. January 16, 2013 Planned greater than 24 hour out of service: 1. December 14, 2012 2. September 3, 2014 3. September 30, 2014 The unplanned non-functional conditions of the seismic monitor have been corrected and were entered into the NSPM Corrective Action Program. 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. Corrective actions are in progress to address the missed reporting of seismic monitor unavailability. The licensee notified the NRC Resident Inspector. | |
| ENS 50870 | 7 March 2015 17:55:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Containment Declared Inoperable Due to Containment Fan Coil Leak | At 1155 CST on March 7, 2015, a small cooling water leak was identified on the 21 Containment Fan Coil Unit east face u-bend on the north east corner bottom bundle. Unit 2 Containment was declared inoperable, which required entry into Technical Specifications (TS) LCO 3.6.1, Condition A, Containment inoperable, applicable in MODES 1, 2, 3, and 4. Immediate actions were taken to isolate the Fan Coil Unit within 1 hour from the initial identification of the leak. 21 Containment Fan Coil Unit was isolated, Containment was declared operable and TS 3.6.1 Condition A was exited at 1220 CST on 3/7/15. This condition is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. The plant remains in a safe condition and there was no effect to the health and safety of the public. The licensee has notified the NRC Resident Inspector. | 05000306/LER-2015-001 |
| ENS 50866 | 5 March 2015 10:06:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unit 2 Declares Unusual Event Due to Inability to Validate a Containment Fire Alarm within 15 Minutes | The Instrument Air Containment Isolation failed closed on Unit 2. This isolated normal letdown / excess letdown and required Pressurizer level to be maintained by diverting Pressurizer level to the Pressurizer Relief Tank. The Pressurizer Relief Tank rupture disc ruptured which resulted in a fire alarm in Unit 2 Containment. The fire alarm could not be validated within 15 minutes which resulted in a declaration of an Unusual Event based on EAL HU2.1. The loss of Instrument Air to Unit 2 Containment resulted in a loss of cooling to the reactor vessel gap and support cooling systems. Due to the loss of reactor vessel ventilation systems a plant shutdown to Mode 3 has been initiated. No radioactive releases to the environment are in progress or expected to occur. The public health and safety has not been jeopardized. The licensee informed state/local agencies and the NRC Resident Inspector and does plan to issue a press release. Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA).
At 1725 (CST) on 3/5/15 Instrument Air to Unit 2 containment was established. Letdown was then reestablished. These actions stabilized the plant and stopped the 30 GPM (gallons per minute) identified leakage out of the PZR Relief Tank rupture disc into containment. This condition is what caused the fire detection alarm in containment, and is also Unusual Event criteria. Unit 2 containment was entered and it was confirmed that NO fire existed in containment. The Unusual Event was terminated at 0018 CST on 3/6/15 based on no fire and no identified RCS leakage into containment. The plant is currently in Mode 3. The health and safety of the public was not jeopardized. The licensee informed state/local agencies and the NRC Resident Inspector and does plan to issue a press release. Notified R3DO (Stone), EO (Ross-Lee) and IRD (Stapleton). Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA). | |
| ENS 50826 | 16 February 2015 22:44:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Special Vent Boundary Inoperable | On February 16, 2015, at 1644 (CST), Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and Unit 2 Control Room Envelope Boundary was declared inoperable when it was discovered that Door 158, Auxiliary Building to 122 Control Room Chiller Room, would not latch. Both trains of Control Room Special Ventilation were declared inoperable and Tech Spec LCO 3.7.10 Condition B was entered. In addition, with the Control Room Envelope Boundary inoperable, Tech Spec 3.7.11 Condition E was required to be entered due to both Control Room Chillers inoperable. The required actions of Tech Spec 3.7.11 Condition E required both Units to enter Tech Spec LCO 3.0.3. As a mitigating action, station personnel were dispatched to secure Door 158. This condition was corrected on February 16, 2015, at 1709 (CST) when the deadbolt was engaged to maintain Door 158 closed. Tech Specs 3.7.10 Condition B, 3.7.11 Condition E, and 3.0.3 were all exited at 1709 (CST) February 16, 2015. This condition is reportable under 10 CFR 50.72 (b)(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function to Mitigate the Consequences of an Accident. Based on immediate implementation of mitigating actions and restoration of the Control Room Envelope, the protection of the health and safety of the public was not affected by this issue. This event has been entered into the sites Corrective Action Program. The licensee has notified the NRC Senior Resident Inspector.
The licensee is retracting this report based on the following: After discussions and interviews with personnel involved, it was determined that the Aux Building Operators found Door 158 latched. After passing through the door, the Operators checked to ensure the door was latched and discovered it was sticking and required assistance to latch by agitating the latch operating mechanism. The door was checked additional times and the door would latch with assistance. The Operators ensured the door was latched and notified the Control Room at 1644 (CST). When the Operators installed the dead bolt and padlock at 1709 (CST), the door was still latched. As the door was never left unlatched and was always able to latch, the door was operable and there was no loss of safety function. The licensee has notified the NRC Resident Inspector. Notified R3DO (Orlikowski). | |
| ENS 50808 | 11 February 2015 04:05:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Containment Declared Inoperable Due to Containment Fan Coil Leak | At 2205 CST on February 10, 2015, a cooling water leak of approximately 60 to 90 drops per minute was identified on the 14 Containment Fan Coil Unit Cooling Water face gasket. As a result, Unit 1 Containment was declared inoperable. This required entry into Technical Specifications (TS) LCO 3.6.1 Condition A, Containment inoperable, applicable in MODES 1, 2, 3, and 4. Immediate action was taken to isolate the fan coil unit within 1 hour from the initial identification of the leak. After isolating the cooling water leak to 14 Containment Fan Coil Unit, containment was declared operable and TS 3.6.1 Condition A was exited at 2232 CST. A Work Request (WR) has been initiated to restore 14 Containment Fan Coil Unit to an operable condition. This condition is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. The plant remains in a safe condition and there was no effect to the health and safety of the public. The licensee has notified the NRC Resident Inspector. | |
| ENS 50801 | 10 February 2015 08:50:00 | 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 | Voids Identified That Result in an Unanalyzed Condition | On February 10, 2015, Prairie Island Unit 1 was shutdown in Mode 3 during a planned outage. Ultrasonic testing in support of Unit 1 Emergency Core Cooling System (ECCS) void verifications identified existing voids with calculated volumes in excess of the OPERABILITY limits specified by the procedure. This rendered both trains of Residual Heat Removal (RHR) systems inoperable requiring entry into Technical Specification 3.0.3 at 0250 (CST). The station took prompt actions to vent the identified voids. The void at 1 RH-12 was vented to within acceptable limits allowing LCO 3.0.3 to be exited at 0538 on February 10. Venting at 1RH-11 is in progress. Voiding was identified at location 1-RH-11 with a calculated volume of 62.21 cubic inches with an OPERABILITY limit of 11.62 cubic inches. Voiding was identified at location 1-RH-12 with a calculated volume of 350 cubic inches with an OPERABILITY limit of 22.84 cubic inches. There was no impact to the health and safety of the public as Safety Injection was available and the time both trains of RHR were INOPERABLE was limited. This 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). The licensee notified the NRC Resident Inspector.
Further analysis was performed on the two void locations, 1RH-11 and 1RH-12. Based on this additional analysis from AREVA, it was determined that the void located at 1RH-11 (RHR Train A) was operable. The calculations for past operability at inspection location 1RH-11 provides reasonable assurance that a void of 65 cubic inches will not generate forces that will fault any piping and supports. The void location at 1RH-12 (RHR Train B) was considered inoperable due to exceeding current procedural operability limits. The void located at 1RH-11 was determined to be nonconforming due to exceeding procedural design basis limits. Therefore, with RHR Train A determined to be operable, this event was not an 8-hour notification for an unanalyzed condition that significantly degrades plant safety, nor 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). The licensee has notified the NRC Resident Inspector. Notified the R3DO (Skokowski). | |
| ENS 50672 | 12 December 2014 08:48:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Auxiliary Building Normal Ventilation Radiation Monitor Failed | At 0248 CST on December 12, 2014, 2R-30, Auxiliary Building Normal Vent Radiation Monitor, failed. This monitor was the redundant monitor for 2R-37, Auxiliary Building Normal Vent Radiation Monitor, which was previously taken out of service for planned maintenance. With 2R-30 and 2R-37 out of service, there are no monitors that will allow for timely classification of two Emergency Action Levels (EALS) - NUE (Notification of Unusual Event) and Alert Classifications. This results in a loss of emergency assessment capability while 2R-30 and 2R-37 are out of service. This is a reportable condition in accordance with 10CFR50.72(b)(3)(xiii). The Auxiliary Building ventilation effluent monitor readings were normal and there were no elevated readings in Auxiliary Building area monitors prior to the unavailability of these monitors. The health and safety of the public was not affected by this issue. The activity to repair and return one of the monitors to service is continuous until restored. The licensee has notified the NRC Resident Inspector. |