Semantic search
| Start date | Title | Description | Topic | |
|---|---|---|---|---|
| ENS 56266 | 9 December 2022 04:01:00 | 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 55529 | 18 October 2021 00:30:00 | 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 53853 | 31 January 2019 13:43:00 | 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 53546 | 10 August 2018 05:00:00 | 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 52638 | 20 March 2017 21:39:00 | 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). | Unanalyzed Condition Past operability |
| ENS 51498 | 26 October 2015 12:03:00 | 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). | Unanalyzed Condition Fire Watch |
| ENS 51429 | 28 September 2015 18:27:00 | 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 50801 | 10 February 2015 08:50:00 | 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). | Unanalyzed Condition Past operability |
| ENS 50523 | 10 October 2014 22:17:00 | Missing Fire Barrier | During a Fire Penetration walkdown, an opening less than 1 (inch) tall under a duct was identified between the Train B Aux Feedwater Pump (AFWP) Room (Unit 1 side) and the Bus150/160 room. This constitutes a missing fire barrier between Fire Area (FA) 32 and FA 37 such that the required degree of separation for redundant safe shutdown trains is lacking. A firewatch has been established on the Unit 1 side of the AFWP Room and Bus 150/160 room. The compensatory fire watch will remain in place per F5 Appendix K until the fire barrier is returned to full functional status. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Resident Inspector of this event.
Northern States Power Company (NSPM) has determined that the opening under the duct is not separating redundant equipment and no longer meets the regulatory definition of an unanalyzed condition. This was identified when it was determined that the redundant pressurizer heater cables credited for safe shutdown were routed through the same fire area. This was reported under EN 50531 (Unanalyzed Condition Due To Missing Fire Barrier). Since the missing fire barrier does not separate redundant equipment, the condition does not degrade plant safety, and is no longer reportable. The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Resident Inspector. Notified R3DO (Peterson). | Safe Shutdown Unanalyzed Condition Fire Barrier Fire Watch |
| ENS 50492 | 24 September 2014 20:33:00 | Unit 1 Train B Rvlis and Iccm Inoperable Due to Data Errors | At 1533 CDT on September 24, during a Past Operability review for the Unit 1 Train B Inadequate Core Cooling Monitor (ICCM), it was discovered that due to data errors, the as left VDC (voltage DC) value for the Remote Display Power Supply was outside of acceptance criteria during the last performance of SP 1213B 'Inadequate Core Cooling Monitor Calibration,' completed on 8/25/2014. The Reactor Vessel Level Instrumentation System (RVLIS) and the Core Exit Thermocouples (CETs) are subsystems of the ICCM System and were determined to be inoperable during this timeframe. Therefore Tech Spec 3.3.3 Condition A for Train B RVLIS and Tech Spec 3.3.3 Condition B for CETs were not met. Also during this time, Unit 1 Train A ICCM was out of service from 1110 (CDT) on 9/8/2014 to 1530 (CDT) on 9/12/2014 for calibration. Thus, both trains of Core Exit Thermocouples and RVLIS were inoperable during this 4 day, 4 hour and 20 minute time period. ICCM System indication is used for Safety Injection Termination and Reinitiation criteria in the Emergency Operating Procedures (EOPs). During the time when both trains of ICCM were inoperable, B Train remained in service and was thought to be operable. This provided the potential for operators to perform an untimely SI Termination or Reinitiation during an accident based on inaccurate information. The Emergency Response Computer System was available to provide accurate backup information to the operators and for timely Emergency Action Limit classification This 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 new information currently being received from Westinghouse, the ICCM Remote Display Power Supply only functions to communicate data to the display and has no effect on the data values. Thus the ICCM System may have been functioning properly. An update to this information will be provided once it is available The protection of the health and safety of the public was not affected by this issue. The Unit 1 Train B ICCM System was returned to operable status at 2300 (CDT) on 9/24/2014. The license has notified the NRC Senior Resident Inspector.
Based on further analysis and input from Westinghouse, the remote display 12V power supply is operable as long as data is still updating on the display. It was determined that the tolerance acceptance criteria within our procedure was conservative. For this application, the power supply can function at much less than 12V. The acceptance criteria within surveillance procedures were incorrect and procedure changes have been submitted to correct the tolerance. Therefore, there was no loss of safety function. The licensee has notified the NRC Resident Inspector. Notified R3DO (Orlikowski). | Past operability |
| ENS 50004 | 7 April 2014 11:00:00 | Control Room Special Vent Boundary Inoperable | At approximately 0600 CDT on April 7, 2014, Units 1 and 2 Control Room Special Vent System (CRSVS) Boundary was declared inoperable when a Control Room door handle fell off. Air could be felt flowing through the hole, causing an opening in the Control Room Envelope (CRE) Boundary. An opening reduces the protection of the Control Room Envelope provided to the operators in the event of a Design Basis Accident. The CRSVS Boundary was declared Inoperable, which required entry into Technical Specifications (TS) LCO 3.7.10. Condition B, one or more CRSVS trains inoperable, in MODES 1, 2, 3, or 4. Actions were immediately initiated to implement mitigating actions. A Work Request (WR) was initiated to repair the door, and work is in progress to restore the Control Room Boundary to an operable condition. This condition is reportable under 10CFR50.72(b)(3)(v)(C), 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. The licensee has notified the NRC Resident Inspector."
Engineering evaluation has determined that the door would still fulfill its safety function of maintaining the integrity of the Control Room Envelope while the door handle was broken. The Control Room Special Vent System Boundary remained operable with the door handle off, therefore, there was no loss of safety function. The licensee has notified the NRC Resident Inspector." Notified R3DO (Riemer). | |
| ENS 49507 | 6 November 2013 06:58:00 | R-21 Circulating Water Discharge Radiation Monitor Planned Outage | When transferring power supplies to a non-safety related cooling tower bus for planned outage maintenance, R-21, the Circulating Water Discharge Radiation Monitor was removed from service at 0058 (CST) and returned to service at 0111 (CST). There is no installed backup for R-21 which has an emergency response function to provide indication of gaseous liquid effluent release to the environment. This monitor has no compensatory measure that will allow timely classification of two NUE (Notification of Unusual Event) and Alert classifications when out of service. This resulted in a loss of emergency assessment capability while R-21 was out of service. There are no radioactive leaks that impact the Circulating Water System. The licensee notified the NRC Resident Inspector. * * * RETRACTION AT 1155 EST ON 12/2/13 FROM WAYNE EPPEN TO DANIEL MILLS * * * Based on further reviews of plant drawings and discussion with the Radiation Monitor System Engineer, R-21 was not inoperable when power supplies were transferred to the non-safety related Cooling Tower Bus. While R-21 was logged out of service during the transfer on November 6, 2013, it did not lose power and was not out of service. R-21 did not lose the ability to provide continuous monitoring of discharge canal effluent or monitoring in the event of an unplanned radiological release. Our defense in depth strategies are relied upon to take actions to protect the health and safety of the public. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Duncan). | |
| ENS 49495 | 1 November 2013 17:06:00 | Both Trains of Auxiliary Building Special Ventilation System Declared Inoperable | During Unit 2 refueling outage (currently defueled) preventative maintenance on CV-31117, Loop B Main Steam Isolation Valve (MSIV), an opening in the valve was discovered without direct administrative controls to ensure the opening could be closed within 6 minutes following a Loss of Coolant Accident on Unit 1. Addition of this opening to other openings created a total of greater than 10 square feet of non-closable openings and required declaring the Auxiliary Building Special Ventilation System (ABSVS) boundary inoperable. The inoperable ABSVS boundary caused both trains of ABSVS to be declared inoperable and required entry into Technical Specification (TS) 3.7.12, Condition B. This could have prevented the ability to control the release of radioactive material and is considered a potential loss of safety function per 10CFR50.72(b)(3)(v)(C). Administrative controls in the ABSVS Boundary were re-established by installing a closure device in the opening on CV-31117 at 1246 CDT. The ABSVS boundary and both trains of ABSVS were declared operable at that time. During the time that the ventilation system was out of service, no evolutions were in progress that could have resulted in an unmonitored release.
After further evaluation, the non-closable boundary openings for the Auxiliary Building Special Ventilation Zone (ABSVZ) including CV-31117, Loop B Main Steam Isolation Valve (MSIV), at the time of discovery is calculated to be 9.48 square feet. This is below the 10 square feet required in TS 3.7.12, thus both trains of ABSVS were operable and no loss of safety function existed. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Cameron). | Time of Discovery |
| ENS 49128 | 17 June 2013 21:15:00 | Potential Unanalyzed Condition - Carbon Dioxide Fire Suppression System Does Not Meet Design Requirements | At approximately 1615 CDT on June 17, 2013, station personnel identified a potential unanalyzed condition based on the following: The 1998 design calculation for the carbon dioxide fire suppression system protecting the Relay Room is in part based on an unverified leakage rate for the enclosure. Recent testing and subsequent calculations have found the room leakage rate is higher than the leakage rate specified in the 1998 design calculation. As such, more carbon dioxide would be required to meet design requirements. The minimum required level for the system storage tank would not currently provide two shots of carbon dioxide (design requirement) to suppress a fire. The minimum required quantity in the tank would be sufficient to provide one shot of carbon dioxide to suppress a fire; however, since this doesn't meet the design requirement, a continuous fire watch was initiated. The licensee notified the NRC Resident Inspector.
Further evaluation determined that the as-found configuration did not meet the minimum design requirements. However, the CO2 concentration was adequate to suppress a fire within the relay and cable spreading room. The licensee has notified the NRC Resident Inspector. Notified R3DO (Hills). | Unanalyzed Condition Continuous fire watch |
| ENS 48870 | 31 March 2013 04:44:00 | Both Trains of Auxiliary Building Special Ventilation System Declared Inoperable | During testing of the Auxiliary Building Ventilation System it was discovered that the outlet damper of one of the normal make-up fans did not close. This created a non-closable opening in the Auxiliary Bldg Special Vent System (ABSVS) boundary of greater than 10 square feet and required entry into TS (Technical Specification) 3.7.12, Condition B, for two ABSVS trains inoperable due to inoperable ABSVS boundary. This is considered a potential loss of safety function to control the release of material. The licensee has notified the NRC Resident Inspector. * * * RETRACTION FROM MARK LOOSBROCK TO PETE SNYDER ON 5/13/13 AT 1639 EDT * * * This notification is being made to retract Event Notification (EN) #48870, which reported a non-closeable opening in the Auxiliary Building Special Vent System (ABSVS) boundary of greater than 10 square feet. This required entry into TS 3.7.12 Condition B for two ABSVS trains inoperable due to an inoperable ABSVS boundary. This was considered a potential loss of safety function to control the release of radioactive material. Engineering evaluation and analysis identified that there was not a loss of safety function because the ABSVS started on demand during the performance of SP 1172 and that the ABSVS Boundary never exceeded the Tech Spec value of 10 square feet per TS 3.7.12. Review of the work order for CD-34042 validated that the system would have performed its safety function during the failure of CD-34042. In the event of a Design Basis Accident the ABSVS would have received a signal to actuate, the system would have performed as designed and drawn a negative differential pressure in the Aux Bldg within 6 minutes with the identified opening of 9.62 square feet that was observed during the failure of CD-34042. The licensee notified the NRC Resident Inspector. Notified R3DO (Riemer). | |
| ENS 48619 | 21 December 2012 15:00:00 | Both Unit Two Auxiliary Feedwater Pumps Declared Inoperable | 21 Motor Driven and 22 Turbine Driven Auxiliary Feedwater Pumps (AFWP) were declared inoperable at 0900 CST on 12/21/2012 due to the Condensate Storage Tank (CST) temperature exceeding 92 degrees F. In accordance with procedure C28.6, 'Condensate Storage Tank Freeze Protection System', the maximum CST temperature shall not exceed 92 degrees F. This is to ensure the maximum AFWP discharge temperature is less than 100 degrees F when an AFWP is at design flow per USAR Table 11.9-2, 'Summary of Assumptions,' used in the AFW system design verification analyses. LCO 3.7.5 Condition D was entered for two AFW trains inoperable in Modes 1, 2 or 3. The AFWP's could start and run if required at the time of entry. Immediate action was taken to reduce the CST temperature. At 1315 CST temperature were lowered below 92 degrees F and LCO 3.7.5 condition D was exited. This condition is reportable per 10 CFR 50.72(b) (3) (v) (D) as an event or condition that could have prevented the fulfillment of a safety function. The licensee has notified the NRC Resident Inspector.
This notification is being made to retract Event Notification (EN) #48619, which reported 21 Motor Driven and 22 Turbine Driven Auxiliary Feedwater Pumps (AFWP) were declared inoperable due to the Condensate Storage Tank (CST) temperature exceeding 92 degrees F. Based on engineering analysis the 92 degree F was a margin value with no formal basis or evaluation. This value was added to the C28.6 procedure in the early 1990's as a precaution to provide Operations with additional operating information. After the 22 CST tank was found above the 92 degrees F temperature on 12/21/12, a formal engineering evaluation (EC 21354) was performed to determine CST heat values and provide an accurate number for temperature margin. The purpose of the CST tank temperature being at a specific value is to ensure that the AFWP discharge temperature stays at or under 100 degrees F to ensure the AF system can provide adequate decay heat removal if called upon during an accident. Engineering evaluation EC 21354 determined that the average CST tank temperatures could reach 96.5 degrees F before the AFWP discharge temperature could have reached 100 degrees F. Therefore, there was no loss of safety function or past operability concerns. The NRC Resident Inspector has been informed. Notified R3DO (Bloomer). | Past operability |
| ENS 48063 | 2 July 2012 20:51:00 | Both Diesel Generators Declared Inoperable | D1 and D2 Diesel Generators (DG) were declared inoperable at 1551 CDT due to exceeding the maximum outside air temperature limit of 97 degree F. This limit was based on heatup analysis for a HELB (High Energy Line Break) in the turbine building. The temperature is measured using the 15 min average of air temperature at the 10 meter meteorological tower, which approximates the intake height for room ventilation. LCO 3.8.1 Condition B was entered for both DGs and LCO 3.8.1 Condition E was entered for two DGs inoperable on the same unit. The DGs could start and run if required at the time of entry. This condition is reportable per 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function. During review of the analysis, two pressure switches on D2 and one on D1 were identified as being limiting and a modification was approved to replace them. Upon replacement, the limiting outside temperature would increase to 100.5 degree F. Replacement of components had been completed for D2 diesel generator prior to LCO entry; however, the temperature limit of 100.5 degree F had not yet been approved. Shortly after D1 was declared inoperable, it was isolated to complete the component replacement which is expected to take less than four hours. D1 will return to operable once the component replacement and PMT is completed. At 1630 (CDT) D2 was declared operable based on approval of a revised operability recommendation that raised the outside temperature limit to 100.5 degree F. Outside air temperature peaked at 97.1 degree F and is slowly decreasing. The licensee has notified the NRC Resident Inspector.
Event Notification 48063 reported that D1 and D2 Diesel Generators (DG) were inoperable due to exceeding the maximum outside air temperature limit of 97 degree F. An additional approved evaluation determined that the D2 lube oil pressure switches were replaced prior to exceeding the outside ambient air temperature limit. The new lube oil pressure switches have a higher temperature rating that supports a higher outside ambient air temperature, therefore, D2 was able to fulfill its safety function and Event Notification 48063 is retracted. The NRC Resident Inspector has been informed. Notified R3DO (Cameron). | |
| ENS 47522 | 9 December 2011 17:59:00 | Unanalyzed Condition- Degraded Fire Barrier Between Fire Areas | A degraded fire barrier between Fire Area (FA) 118 (Bus 26 Room) and FA 128 (Bus 27 Room) has existed during the last three years. The top of the wall between the Bus 26 and Bus 27 room had a missing/degraded fire barrier. At the time of discovery on December 9, 2011 a fire watch was in place. However, it was determined that at various times during the last three years a fire watch was not established. Bus 27 has been conservatively aligned to Bus 26 to provide the required degree of separation for redundant safe shutdown trains (between Bus 26 and Bus 25). The fire watch will remain in place as a compensatory measure until the fire barrier is repaired. (The) NRC Resident (Inspector) has been informed.
An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was reported on December 13, 2011 for a degraded fire barrier between Fire Area (FA) 118 (Bus 26 Room) and FA 128 (Bus 27 Room.) 'Subsequent engineering analysis determined that the degraded fire barrier maintained the required degree of separation for redundant safe shutdown trains and plant safety was not significantly degraded. The 10 CFR 50.72(b)(3)(ii)(B) report is retracted. The NRC Resident Inspector has been informed. Notified R3DO (L. Kozak) | Safe Shutdown Time of Discovery Fire Barrier Fire Watch |
| ENS 47321 | 5 October 2011 12:27:00 | Two Loops of Reactor Coolant System (Rcs) Were Declared Inoperable | Unit 2 was shut down at 1048 CDT on October 4, 2011 due to 22 Reactor Coolant Pump (RCP) low seal leak off flow. At 0727 CDT on October 5, 2011 with Unit 2 in Mode 4 and 22 RCP secured, 21 RCP was also secured due to low seal leak off flow. It was determined that the station would not consider RHR (residual heat removal) available for decay heat removal per Technical Specification (TS) due to a procedural issue related to component cooling shutdown lineups. RCS temperature is being maintained stable in Mode 4 between 320 and 340 degrees Fahrenheit using natural circulation and the steam dump to the condenser. No operations were permitted that would have caused introduction into the RCS, coolant with boron concentration less than required to meet the Shutdown Margin. At approximately 1329 CDT, 21 RCP was restarted after the flow transmitters were vented and 21 RCP seal leak off flow was determined to be acceptable. The RHR procedural issue has been resolved. Per TS LCO 3.4.6, 'Two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be OPERABLE, and one loop shall be in operation.' This condition is reportable as an event or condition that could have prevented fulfillment of a safety function under 10CFR 50.72(b)(3)(v)(B). During this event, RHR could not be credited because the utility was not able to valve in component cooling water to a SFP heat exchanger due to procedural issues. With the restart of 21 RCP and shutdown cooling water alignment, TS 3.6.4 A and B statements have been met and exited. 21 RCP loop in operation and 22 RHR is operable and can be aligned for shutdown cooling. Both Unit 2 EDG's are available and offsite power is normal electrical lineup. The licensee is no longer in any TS LCO statements. The NRC Resident Inspector has been notified.
An eight hour report (EN #47321) per 10CFR 50.72(b)(3)(v)(B) was conservatively reported because both trains of Residual Heat Removal (RHR) system and both trains of Reactor Coolant System (RCS) were thought to be unavailable. Additional review indicated that 21 RCP and 22 RCP were undamaged, functional, and available for loop operation; therefore, there was no loss of safety function as RCPs were available. The station evaluated 21 and 22 RCP seal leak off flow performance and subsequently restarted 21 RCP and restored forced cooling. The 21 RCP seal leak off flow has remained stable in the acceptable range. 21 RCP was available throughout this evolution. If needed in Mode 4, 22 RCP was also available. The procedures that were in place and available to the operators prior to 1320 CDT on October 5, 2011, would have allowed RHR operation; however, to supply RHR to Unit 2, Unit 1 would have entered an LCO (for supplying Component Cooling (CC) water to the Spent Fuel Pool (SFP) heat exchangers) to maintain RHR's safety function. After 1320 CDT, procedure revisions allowed operation of RHR without an associated Unit 1 LCO entry. RHR remained operable throughout this evolution. In summary, a loss of safety function for both trains of RHR and RCS did not exist and the 50.72(b)(3)(v)(B) report (EN # 47321) is retracted. The NRC Resident Inspector has been informed. R3DO(Phillips) notified. | Shutdown Margin |
| ENS 46934 | 7 June 2011 18:49:00 | Both Emergency Diesel Generators Declared Inoperable Due to Excess Outside Ambient Air Temperature | Outside ambient air temperature exceeded the maximum analytical value for operability for Unit 1 D1 and D2 Diesel Generators at 1349 CDT. The calculated limiting outside air temperature needed for equipment in the D1 and D2 rooms to meet their temperature limits is 100.5?F. Outside ambient temperature exceeded this limiting value and both Unit 1 safeguards diesel generators were declared inoperable at 1349 CDT on 6/7/2011. If outside ambient air temperature is above the maximum analytical value, components within the D1 and D2 diesel rooms may not be able to perform their required functions thus preventing them from fulfilling their safety function needed to mitigate the consequences of an accident (10 CFR 50.72 (b)(3)(v)(D)). Unit 1 is currently in Mode 3, Hot Standby. Ambient outside air temperatures are at or near peak values for the day and expected to decrease approximately 1 to 2 degrees per hour which will restore ambient conditions to less than the maximum analytical value. The NRC Resident Inspector has been notified. The outside air temperature has peaked at 101.4?F which is unusually high for this location and is expected to drop below the 100.5?F limit shortly. The licensee does not anticipate that this condition will be repeated again any time soon.
An eight hour report (EN #46934) per 10 CFR 50.72(b)(3)(v)(D) was made on June 7, 2011 for outside ambient air temperature exceeding the maximum analytical value for operability for Unit 1 D1 and D2 Diesel Generators. Subsequent engineering analysis determined that D1 & D2 diesel generators would have performed their safety function with outside air temperatures of up to 102.5 degrees F. The outside air temperature did not reach 102.5 degrees F and the diesel generators were not inoperable. The 10 CFR 50.72(b)(3)(v)(D) report, EN #46934, is retracted. The NRC Resident Inspector has been notified. Notified R3DO (Giessner). | |
| ENS 46699 | 25 March 2011 15:53:00 | Potential Unanalyzed Condition Due to Power Level Greater than Limit | During performance of maintenance to troubleshoot the B feedwater regulating bypass valve, the Thermal Power Monitor (TPM) indication exceeded the maximum thermal power assumed in the Safety Analysis Report. Operators were maintaining 12 Steam Generator (SG) water level in a band from 40 to 48 percent by controlling the B Feed Regulating Valve (FRV) in manual from the Control Room. Operators noted a power increase; adjustments were made via the FRV to reduce SG water level, however the valve response was sluggish and thermal power exceeded 100%. Immediate steps were taken to reduce power to below 100% by reducing 1st stage turbine pressure and inserting Bank D control rods 7 steps. The TPM indication was above the maximum thermal power limit of 100.36% for 1.68 minutes. The TPM indication peak was 100.39%. No concurrent increase in power was observed by the nuclear indication system. NRC Resident had been informed.
An eight hour report (EN #46699) per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on March 25, 2011 for Thermal Power Monitor indication above the maximum thermal power limit of 100.36% for 1.68 minutes. Subsequent engineering investigation has determined that this specific transient had been previously analyzed. The transient was within the bounds of the safety analysis. The 10 CFR 50.72(b)(3)(ii)(8) report (EN #46699) is retracted. NRC Resident has been informed. Notified R3DO (Peterson). | Unanalyzed Condition |
| ENS 46318 | 7 October 2010 18:52:00 | Degraded Fire Barrier | During a walkdown for the Fire Penetration Seal Project a degraded fire barrier was identified in the wall between the Unit 1 and Unit 2 Normal 480V Switchgear Rooms. The wall is listed as an Appendix R wall between Fire Area (FA) 37 and FA 38. The wall separates redundant safe shutdown cables. There is a 2 inch gap where the top of the wall meets the ceiling. The gap is filled with combustible foam - some loose and some stationary. This has been identified as a missing fire barrier such that the required degree of separation for redundant safe shutdown trains is lacking. A fire watch has been established as a compensatory measure and will remain in place until the fire barrier is repaired. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). The licensee has notified the NRC Resident Inspector of this event.
An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on October 7, 2010 for a degraded fire barrier between Fire Area (FA) 37 and FA 38. Subsequent engineering analysis determined that the degraded fire barrier maintained the required degree of separation for redundant safe shutdown trains and plant safety was not significantly degraded. The 10 CFR 50.72(b)(3)(ii)(B) report is retracted. The NRC Resident (Inspector) has been informed. Notified the R3DO (Cameron). | Safe Shutdown Unanalyzed Condition Fire Barrier Fire Watch |
| ENS 46261 | 16 September 2010 18:43:00 | Unanalyzed Condition - Degraded Fire Barrier Identified | This is a late eight hour report submitted under 10CFR50.72(b)(3)(ii), 'Degraded or Unanalyzed Condition.' During walk downs for the Fire Penetration Seal Project on September 16, 2010 at 1343 (hrs. CDT), a degraded fire barrier was identified in the wall between the Unit 1 and Unit 2 sides of the Electrical Piping Area. The wall is listed as an Appendix R wall between Fire Area (FA) 29 and FA 30. The wall separates safety related redundant cables. There is an approximately one inch gap filled with loose fitting foam from the top of the wall to the concrete ceiling above it. In one place the foam has a gap approximately one inch wide. This has been identified as a missing fire barrier such that the required degree of separation for redundant safe shutdown trains is lacking. A fire watch was established as a compensatory measure on 9/16/10. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10 CFR 50.72(b)(3)(ii)(B). The licensee has notified the NRC Resident Inspector of this event. The fire watch remains in place.
An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on September 20, 2010 for a degraded fire barrier between Fire Area (FA) 29 and FA 30. Subsequent engineering analysis determined that the degraded fire barrier maintained the required degree of separation for redundant safe shutdown trains and plant safety was not significantly degraded. The 10 CFR 50.72(b)(3)(ii)(B) report is retracted. The NRC Resident (Inspector) has been informed. Notified the R3DO (Cameron). | Safe Shutdown Unanalyzed Condition Fire Barrier Fire Watch |
| ENS 45855 | 19 April 2010 03:25:00 | Lco 3.0.3 Entry and Loss of Safety Function Due to Loss of Turbine Building High Energy Line Break Compensatory Measure | At 2225 CDT on 4/18/2010, Operations discovered that the Unit 1 Turbine Building Truck Aisle Rollup Door Security Fence was closed. This fence was to be maintained open as the truck aisle is a required drainage path from the Unit 1 Turbine Building to outside in the event of flooding resulting from a High Energy Line Break (HELB). With the expanded metal mesh door (fence) closed, turbine building debris could clog the drainage path and result in a higher than calculated water level being reached for this event. As the final water level cannot be predicted, this represented an unanalyzed condition. The higher water levels would be reached at least one hour after the postulated turbine building HELB event. High water levels could result in a Loss of Safety Function for Unit 1 Emergency Diesel Generators. Auxiliary Feedwater (both units) and DC Electrical Power (both units) if water levels exceed critical heights in the associated rooms. The doors (fence) were reopened at 2227 CDT. Unit 1 entered LCO 3.0.3 for this two minute period. With the doors (fence) opened, Unit 1 and 2 LCO conditions were again satisfied. The Unit 2 truck aisle was in the assumed condition, and Unit 2 is in Mode 5 so loss of AFW or DC power would not result in an LCO 3.0.3 entry. Unit 2 Emergency Diesel Generators are not affected as the Unit 2 truck aisle drain path would prevent water levels from reaching critical heights. The initial investigation determined that a Security Officer had closed the gate at approximately 1855 CDT on 4/18/2010. The NRC Resident Inspector has been notified. A press release is not planned.
In reference to the 4/18/2010 discovery of the security fences (screen door) closed on the Unit One Turbine Building truck aisle door and the Tech Spec LCO 3.0.3 entry on Unit One, further analysis of HELB flooding scenarios in the plant configuration known to be in effect at that time has shown that the LCO 3.0.3 entry was not warranted. Engineering analysis demonstrates that water levels in the Unit One Turbine Building from a postulated worst case HELB flooding event would only have rendered the 12 Motor Driven Auxiliary Feedwater Pump inoperable. Thus TS LCO 3.7.5 Condition B would have been the only LCO action statement required to be entered. No loss of safety function existed." The licensee will notify the NRC Resident Inspector. Notified the R3DO (Skokowski). | Unanalyzed Condition |
| ENS 43010 | 25 November 2006 01:45:00 | Unanalyzed Condition | With Prairie Island Unit 2 core offloaded for the 2R24 refueling outage, it was discovered that a potential common cause failure of the 21 and 22 Residual Heat Removal (RHR) Pumps may exist. A non-safety related 120 VAC motor heater circuit from Panel 2RPA3 Circuit 28 supplies power to 21 and 22 RHR Pumps. Investigation continues. The RHR system is not required by the current operating mode. The RHR Pumps on the operating unit (Unit 1) are supplied from separate 120 VAC panels so the concern does not exist for the 11 and 12 RHR pumps. The licensee notified the NRC Resident Inspector.
Event Number 43010 (reported 11/24/06) reported a potential single failure vulnerability of the Unit 2 residual heat removal (RHR) pumps. NMC identified that power cables for the motor heaters in each pump were powered from the same 120 Volt panel and that the heater power cables were routed for some distance in the same conduit. The concern was that an electrical fault could propagate from one RHR pump motor to the opposite train RHR pump motor. Further engineering evaluation has determined that no credible failure mechanism exists that could cause a fault in one RHR pump motor to propagate to the opposite train RHR pump motor. Since the dielectric strength of insulation on the heaters is greater than the dielectric strength of the air between the motor windings and the motor frame or the stator, any fault in the motor windings would be to ground (and, thus, interrupted by the ground fault protective relaying). Therefore, no potential single failure vulnerability existed in the as-found condition and NMC hereby retracts the event reported on 11/24/06 (Event Number 43010). The licensee notified the NRC Resident Inspector. Notified the R3DO (H. Peterson). | Unanalyzed Condition |
| ENS 41377 | 6 February 2005 02:46:00 | 4160 Volt Relaying and Metering Single Failure Vulnerability | This report is being made pursuant to 10CFR50.72(b)(3)(ii)(B). At 20:46 on 02/05/05 while reviewing Operation Experience from another Nuclear Site, it was discovered that the design of the AC Auxiliary Power System incorporates a common circuit which could result in a bus lockout preventing the re-energization of both unit one safeguards buses from either onsite or offsite power sources due to a single failure in the common portion of the circuitry. This common circuit includes various metering circuits and is also connected to overcurrent devices that feed breaker lockout relays. Should a failure occur on these common circuits, all breakers supplying power to these buses would be opened, locked out, and prevented from reclosure onto the buses. Due to the loss of the ability to accommodate a single failure, one offsite power source ( CT11 ) has been declared inoperable and Technical Specification 3.8.1 Condition A, entered. Corrective actions are in progress to isolate the common circuit and eliminate the single point vulnerability. The licensee notified the NRC Resident Inspector. See similar events #41362 (Crystal River), #41366 (LaSalle), #41369 (Quad Cities) ,#41370 (Dresden) and #41377 (Monticello). * * * RETRACTION FROM S. SEILHYMER TO S. ROTTON AT 1140 ON 6/13/06 * * * The event reported on February 6, 2005 (4160 volt relaying and metering single failure vulnerability, NRC Event Number 41377) is hereby retracted. Subsequent analysis has concluded that a common mode failure resulting in a simultaneous lockout of both safeguards 4160 volt Bus 15 and Bus 16 due to the common metering circuit was not a credible event in the as-found condition. The analysis showed that the required magnitude of a fire induced fault is much greater than the available fault current in any of the associated circuits. The analysis also showed that dual bus lockout due to an open circuit under non-fire induced conditions is not credible since the imbalance current will be less than the trip setpoint of the affected relays. Thus, although the plant had been in an unanalyzed condition, that condition has now been shown to have not significantly degraded plant safety and is, therefore, not reportable under the requirements of 10 CFR 50.72(b)(3)(ii)(B). Additionally, NMC will be submitting a letter to cancel the associated LER 1-05-01. The licensee notified the NRC Resident Inspector. Notified R3DO (Louden). | Unanalyzed Condition |