Semantic search
| Start date | Title | Description | Topic | |
|---|---|---|---|---|
| ENS 54938 | 8 October 2020 16:25:00 | Notification of an Unusual Event | At 1125 CDT on 10/8/20 Monticello Nuclear Generating Plant declared a Notification of an Unusual Event (NOUE) due to a Security Condition that did not involve a hostile action, due to a helicopter that hovered over the site for approximately 10 minutes. The unit remained at 100 percent power during the event. The licensee notified the NRC Resident Inspector, the Minnesota State Duty Officer and the Wright County and Sherburne County Sherriff departments of the event. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
At 1715 CDT Monticello Nuclear Generating Plant terminated the Notice of Unusual Event (HU1.1), upon confirmation through federal law enforcement and conversation with the aircraft owner that the aircraft in question was performing power line inspections for a different utility and was not a threat to the plant. Monticello determined that this condition did not meet the 1-hour reporting requirements of 10 CFR 73.71. Notified R3DO (Orlikowski), NRR EO (Miller), and IRD MOC (Gott). Additionally, notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
Monticello Nuclear Generating Plant (MNGP) is retracting this event notification based on subsequent information received that was not available at the time of the original notification. Following review of the additional information collected by the site through an investigation performed with input from local and federal law enforcement, Xcel Energy determined the helicopter did not constitute a credible threat or compromise site safety or security. There was no impact to public health or safety. The NRC Resident Inspector has been notified. The licensee will notify the State and county. Notified R3DO (Peterson). | |
| ENS 53997 | 12 April 2019 23:15:00 | En Revision Imported Date 5/28/2019 | EN Revision Text: HIGH ENERGY LINE BREAK DOOR FOUND IN INCORRECT POSITION RESULTING IN LPCI AND CORE SPRAY BEING INOPERABLE At approximately 1815 CDT on April 12, 2019, High Energy Line Break (HELB) Door-410A in the Reactor Building was discovered in the closed position. HELB Door-410B was previously closed for maintenance. Either Door-410A or Door-410B must be open to support the current HELB analyses. With both doors closed, this is considered an unanalyzed condition resulting in the loss of a post-HELB safe shutdown path. With Door-410A and Door-410B closed, LPCI (Low Pressure Coolant Injection) and Core Spray injection valves in both divisions are no longer considered available. This condition is being reported under 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function. The condition was resolved at approximately 1845 CDT on April 12, 2019 when Door-410A was blocked open. The health and safety of the public was not affected by this condition. The NRC Resident has been notified.
Event Notification (EN) #53997, made on 4/13/2019, is being retracted. An engineering evaluation completed subsequent to this event analyzed the discovered condition with both Door-410A and Door-410B being closed. The engineering evaluation determined that the environmental conditions present with both Door-410A and Door-410B closed would not have impacted the availability of both divisions of the LPCI (Low Pressure Coolant Injection) and Core Spray injection valves nor would it have resulted in the loss of a post-HELB safe shutdown path. Therefore, this condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degrades plant safety or 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 NRC Resident Inspector has been notified. The licensee also notified the Minnesota State Duty Officer. Notified R3DO (Cameron). | Safe Shutdown Unanalyzed Condition |
| ENS 52454 | 21 December 2016 15:35:00 | High Pressure Coolant Injection System Inoperable | At 0935 (CST) on 12/21/2016, while performing the High Pressure Coolant Injection (HPCI) Comprehensive Pump and Valve Tests for post-maintenance testing following scheduled maintenance, the HPCI turbine did not start as expected due to the HPCI turbine stop valve failing to open. This issue is being reported under 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. Investigation into the failure of the HPCI system to start is in progress. The plant remains at 100% power with no challenges to the health and safety of the public. The NRC Resident Inspector has been notified. The plant is in a 14-day action statement under LCO 3.5.1, 'ECCS - Operating' due to the HPCI turbine stop valve failure. The licensee notified the Minnesota State Duty Officer.
On December 21, 2016, the NRC Operations Center was notified of Event Number 52454 that described a failure of the High Pressure Coolant Injection (HPCI) turbine stop valve to open during post maintenance testing prior to being declared operable. The condition was reported in accordance with 10 CFR 50.72 (b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. At the time, it was not readily apparent that the failure was due to the maintenance activities. Subsequent return-to-service testing showed the oil system vent and fill had been inadequate following the maintenance. This event occurred as a result of the maintenance process and would not have occurred during normal operation of the system. NUREG-1022, Revision 3 states, 'reports are not required when systems are declared inoperable as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that would have resulted in the system being declared inoperable).' There was no discovered condition that would have resulted in the safety function of the system being declared inoperable under normal, non-maintenance conditions. Based on the above additional information, Monticello Nuclear Generating Plant is retracting this report. The plant was in a planned evolution and did not discover a condition that could have prevented performing a safety function. The licensee has notified the NRC Resident Inspector. Notified R3DO (McCraw). | Time of Discovery |
| ENS 50915 | 21 March 2015 10:37:00 | High Pressure Coolant Injection Inoperable Due to Condensation in Steam Line | At 0537 CDT on March 21, 2015, following the High Pressure Coolant Injection (HPCI) system quarterly pump and valve surveillance, after HPCI was removed from service, an alarm for the HPCI Turbine Inlet High Drain Pot Level did not reset. This indicated that LS-23-90 (HPCI Steam Supply Drain High Level Bypass) did not reset, which could be an indication that condensate exists in the steam line. The system responded as designed but the alarm did not clear as expected. Without assurance that the condensate has been removed from the HPCI steam line, HPCI remains inoperable for reasons other than the planned surveillance. As a result, this condition is being reported under 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented fulfillment of the safety function at the time of discovery. The health and safety of the public was maintained as the plant was in a normal condition with no initiating event in progress. The NRC Resident Inspector has been notified. The State of Minnesota will be notified.
On March 21, 2015, Northern States Power Minnesota reported a condition that could have prevented the fulfillment of a safety function under 10 CFR 50.72(b)(3)(v)(D). The High Pressure Coolant Injection (HPCI) System was declared inoperable for a reason other than planned maintenance due to the failure of the HPCI Steam Supply Drain Hi Level Bypass Level Switch to clear the high level alarm subsequent to actuation. An engineering evaluation was performed and concluded that the function of the primary pathway to remove condensate remained unchallenged by the condition present on the level switch This conclusion was also validated via thermography with the HPCI steam supply pressurized and bypass valve open. The verification that the primary pathway was functional provides reasonable assurance that the HPCI steam supply was always clear of condensate supporting the ability of HPCI to perform its required safety function. Therefore, the condition present on the level switch did not render HPCI inoperable. The conclusions of the engineering evaluation provide the basis for retraction of the ENS report made on March 21. The NRC Resident Inspector has been notified. The licensee will also notify the State of Minnesota. Notified R3DO (Peterson). | Time of Discovery |
| ENS 50899 | 16 March 2015 23:20:00 | High Pressure Coolant Inject Declared Inoperable Following Scheduled Maintenance | At 1820 on March 16th, 2015, the High Pressure Coolant Injection (HPCI) system steam lines were re-pressurized following scheduled maintenance. Upon restoration, an alarm was received that indicated condensate may exist in the steam line. The system responded as designed but the alarm did not clear as expected. Without assurance that the condensate has been removed from the HPCI steam line, HPCI remains inoperable for reasons other than the planned maintenance. As a result, this condition is being reported under 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented fulfillment of the safety function at the time of discovery. The health and safety of the public was maintained as the plant was in a normal condition with no initiating event in progress. The NRC Resident Inspector has been notified. The licensee will also notify the State of Minnesota.
On March 16, 2015, Northern States Power Minnesota reported a condition that could have prevented the fulfillment of a safety function under 10 CFR 50.72(b)(3)(v)(D). The High Pressure Coolant Injection (HPCI) System was declared inoperable for a reason other than planned maintenance due to the failure of the HPCI Steam Supply Drain Hi Level Bypass Level Switch to clear the high level alarm subsequent to actuation. An engineering evaluation was performed and concluded that the function of the primary pathway to remove condensate remained unchallenged by the condition present on the level switch. This conclusion was also validated via thermography with the HPCI steam supply pressurized and bypass valve open. The verification that the primary pathway was functional provides reasonable assurance that the HPCI steam supply was always clear of condensate supporting the ability of HPCI to perform its required safety function. Therefore, the condition present on the level switch did not render HPCI inoperable. The conclusions of the engineering evaluation provide the basis for retraction of the ENS report made on March 17. The NRC Resident Inspector has been notified. The licensee will also notify the State of Minnesota. Notified R3DO (Peterson). | Time of Discovery |
| ENS 50806 | 10 February 2015 18:40:00 | Unanalyzed Condition in Station Blackout Implementation at Monticello | On February 10, 2015, at 1240 EST, Northern States Power-Minnesota (NSPM) determined that the Station Blackout (SBO) implementation at Monticello Nuclear Generating Plant (MNGP) was not consistent with the NRC Safety Evaluation (SE). Specifically, the High Pressure Coolant Injection (HPCI) system was not being utilized in a manner consistent with the NRC SE for SBO. Current battery calculations do not reflect a full complement of HPCI system equipment running for the duration (coping requirements) of the SBO event. The calculation assumed a manual action to remove the HPCI auxiliary oil pump from operating during an SBO event in order to preserve the station battery. NSPM is reporting this as an Unanalyzed Condition pursuant to the requirements of 10 CFR 50.72(b)(3)(ii)(B). The health and safety of the public was not affected since no SBO event occurred. All station batteries and the HPCI system remain operable in accordance with the plant Technical Specifications. The NRC Resident Inspector was notified of the event.
An engineering analysis was performed updating the battery calculations for Station Blackout (SBO) implementation demonstrating the ability of the safety related station batteries to provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of the SBO for the specified four hours. Therefore, the battery calculation is analyzed and specifically the High Pressure Cooling Injection (HPCI) System is analyzed to run in automatic for the entire duration of the SBO event meeting the site licensing basis for SBO. The SBO procedure has been revised to incorporate HPCI running in automatic for the entire duration of the SBO event. The NRC Resident Inspector has been notified. Notified R3DO (Duncan) | Unanalyzed Condition |
| ENS 50663 | 10 December 2014 00:30:00 | High Energy Line Break Door Found Closed | At 1830 (CST) on December 9, 2014 Door 410B, a HELB (High Energy Line Break) door between the east and west sides of the ground floor of the reactor building, was found closed. This door is one half of a pair of double doors that are normally open to provide a HELB energy and flooding release path to mitigate postulated HELB events. The closed HELB door has the potential to impact safe shutdown by exposing both divisions of safe shutdown components to unanalyzed environmental conditions. With the potential loss of both divisions of safe shutdown equipment, no safe shutdown path would exist. This condition is being reported as an unanalyzed condition as defined by (10 CFR) 50.72(b)(3)(ii)(B). The HELB door was immediately opened and returned to normal configuration. Door 410A remained open during the time that Door 410B was closed and provided an available, but not yet analyzed, release path that could have mitigated the consequences of this event. The health and safety of the general public was not impacted as a result of this condition. The NRC Resident Inspector has been notified.
Further analysis has determined that the condition did not significantly degrade plant safety. Door 410B in the Reactor Building was found closed. This door is one half of a pair of double door (Doors 410A and 410B) that normally open to provide a High Energy Line Break (HELB) energy and flooding release path to mitigate postulated HELB events. The condition of one half of the double door closed was not previously analyzed. A subsequent completed engineering evaluation analyzed this condition, Door 410B being closed and Door 410A being open, for the following environmental conditions: peak compartment temperatures, block wall differential pressure, radiation dose, and flooding. The environmental conditions found the Reactor Building in response to Door 410B being closed with 410A being open does not affect the operability of safety related equipment housed within the Reactor Building or the ability to safely shut-down the plant and maintain the plant shutdown condition following a HELB event. The NRC Resident Inspector has been notified. Notified R3DO (Dickson). | Safe Shutdown Unanalyzed Condition |
| ENS 50456 | 14 September 2014 07:26:00 | Operation in Unanalyzed Region of the Power to Flow Map | At 0226 CDT on September 14, 2014, MNGP (Monticello Nuclear Generating Station) experienced a trip of the 12 Reactor Recirc Pump. The subsequent power drop and lowering of recirculating water flow resulted in the plant being outside of the analyzed region of the Power to Flow Map. Operators promptly restored operation within the analyzed region per procedural guidance. This event has been determined to be a condition where the plant was in an unanalyzed condition that significantly degrades plant safety and is reportable under 50.72(b)(3)(ii). The plant is in stable condition at 51% power and the health and safety of the public were not affected. The investigation of the cause of this event is in progress. The licensee will notify the NRC Resident Inspector.
Further analysis has determined that the condition did not significantly degrade plant safety. General Electric Hitachi was requested to review the event and confirm the SIL653 guidance remains applicable for MELLLA+ (Maximum Extended Load Line Limit Analysis) operation. This review was completed and the conclusions of SIL653 remain valid . The SIL states that: 'unplanned events that result in the plant exceeding the licensed upper boundary do not constitute a safety concern. The consequences of such unplanned events are bounded by the GE safety analysis of limiting events initiated from within the licensed operating domain. Stability monitoring and protection using Detection and Suppression Solution Confirmation Density remained available throughout the event (oscillating power range monitors). The NRC Resident Inspector has been notified. Notified R3DO (Hills). | Unanalyzed Condition |
| ENS 49363 | 19 September 2013 22:50:00 | High Pressure Coolant Injection Inoperable Due to Steam Leak | On 9/19/2013, during the performance of the High Pressure Coolant Injection (HPCI) quarterly pump and valve surveillance, a steam leak was discovered. HPCI had previously been declared inoperable due to planned maintenance. As a result of the steam leak, HPCI remains inoperable. Action taken: 14 days Required Action TS 3.5.1.J.2 remains in effect and corrective actions are in progress. The licensee has notified the NRC Resident Inspector. * * * RETRACTION FROM RANDY SAND TO PETE SNYDER AT 1546 EDT ON 10/28/13 * * * The licensee performed an evaluation that determined the minor steam leak from the High Pressure Coolant Injection (HPCI) turbine reported on 9/20/2013 was not significant enough to prevent HPCI from mitigating the consequences of an accident or mitigating a Station Blackout (SBO) event. The licensee performed an engineering evaluation of the HPCI system, the HPCI pump/turbine and the HPCI room environmental conditions assuming conservative leakage conditions existed. The results of this evaluation confirmed that the HPCI system would have been able to perform its design function assuming conservative leakage conditions existed throughout limiting events. The HPCI pump/turbine would not have failed during any accident or SBO event, and sufficient motive (steam) force was available for the HPCI system to perform its design functions. There would have been no unacceptable impact on the HPCI pump/turbine oil system due to the steam leak. The HPCI room environment would not have exceeded allowable limits. For events where AC power is available, the analysis took advantage of the HPCI room cooler that is powered from an essential power source and supplied from a safety related service water system. This cooler was available during the period of the steam leak. The evaluation of room conditions for SBO conditions did not include use of the HPCI room cooler and also showed room conditions would have remained within acceptable values. There would not have been a buildup of fluid sufficient to cause a flood in the HPCI room. Therefore, based on the results of the formal engineering evaluation, the HPCI system was capable of performing its safety function and therefore, this event may be retracted. The licensee will notify the NRC Resident Inspector. Notified R3DO (Daley). | |
| ENS 49335 | 9 September 2013 19:00:00 | High Energy Line Break Barrier Improperly Obstructed | At 1400 (CDT) on September 9, 2013, plant personnel found HELB barrier HATCH-1/TB blocked. The hatch is diamond plate steel located on the turbine floor. Pallets, a fan and a gantry were positioned on top of the hatch possibly preventing pressure relief during a HELB event. This issue is being reported as an unanalyzed condition per 10CFR50.72(b)(3)(ii)(B). All items were removed from the hatch by 1800 on September 9, 2013. The licensee notified the NRC Resident Inspector.
The licensee reviewed the design basis calculations and analyses for HELB events in the area where Hatch-1/TB is located. The review determined that it is acceptable to block or hold Hatch-1/TB down and that the painted markings on the hatch are overly restrictive. In looking at HELB calculation of record, feed water break at the feed water pumps, there is no flow path modeled between the HELB volumes. Therefore, if Hatch-1/TB is blocked, the physical flow path is in accordance with the Gothic model of the HELB volume. Further, the analyses also indicated that no credit is taken for the hatch to relieve as no HELBs are postulated in the room under the hatch. Finally, Hatch 1/TB structural integrity was verified assuming a HELB occurred with the hatch blocked as described in notification 49335 (pallets, fan and gantry) the barrier would function as designed. Licensee initiated a Work Request to remove any markings on Hatch-l/TB that indicate do not block or hold down. The licensee will notify the NRC Resident Inspector. The Region 3 Duty Officer (Valos) was notified. | Unanalyzed Condition |
| ENS 49316 | 1 September 2013 21:10:00 | Recirc Pump Runback and Power Reduction | On September 1, 2013 at 1610 CST, Monticello Nuclear Generating Plant (MNGP) experienced a runback of 'A' Recirc pump from 87% speed to 82% speed. Operators took action to lock 'A' Recirc pump scoop tube. This runback resulted in a power reduction from 100% to 98% RTP (Rated Thermal Power). Should a LOCA (Loss of Coolant Accident) occur with the resultant mismatch between total jet pump flows of the two loops greater than required limits, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. It has been determined that this is an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). At 2123 CST on September 1, 2013, MNGP completed adjusting recirc flow speed on 'A' and 'B' Recirc pumps to match jet pump loop flows to within the required limits and is no longer in an unanalyzed condition. Both 'A' and 'B' Recirc scoop tubes remain locked pending investigation. The licensee has notified the NRC Resident Inspector.
This is a retraction for ENS 49316: The licensee reviewed the MNGP design basis analysis to determine if the event was bounded. The licensee determined that the Loss of Coolant Accident (LOCA) provides a bounding analysis for this event. The limiting LOCA event for the MNGP as analyzed in accordance with 10CFR50 Appendix K conditions is based upon single failure of the Low Pressure Coolant Injection (LPCI) injection valve, effectively making LPCI inoperable for the event. The large break Design Basis Accident (DBA) with LPCI injection valve failure (which is analytically equivalent to the condition of both LPCI subsystems being inoperable) is the event analyzed for the current Licensing Basis Peak Cladding Temperature (PCT). This analysis bounds the event as a recirculation pump flow mismatch event is less limiting than the LOCA with LPCI injection valve failure analysis. Therefore, this recirculating loop flow mismatched event is less limiting than a previously analyzed event and ENS 49316 may be retracted as an unanalyzed event. The licensee has notified the NRC Resident Inspector. Notified R3DO (Skokowski). | Unanalyzed Condition |
| ENS 49312 | 28 August 2013 02:52:00 | Recirc Pump Runback and Power Reduction | On August 27, 2013 at 2152 CDT, Monticello Nuclear Generating Plant (MNGP) experienced a runback of (the) 'B' Recirc pump from 87% speed to 71% speed. Operators took action to lock (the) 'B Recirc pump scoop tube. This runback resulted in a power reduction from 100% to 94% RTP (Rated Thermal Power). With the resultant mismatch between total jet pump flows of the two loops greater than required limits, should a LOCA (Loss of Coolant Accident) occur, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. It has been determined that this is an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(8). At 0121 CST on August 28, 2013, MNGP completed reducing power to 88% using 'A' Recirc pump to match total jet pump flows and (the plant) is no longer in an unanalyzed condition. The 'B' Recirc scoop tube remains locked pending investigation. The licensee has notified the NRC Resident Inspector.
This is a retraction for ENS 49312: The licensee reviewed the MNGP design basis analysis to determine if the event was bounded. The licensee determined that the Loss of Coolant Accident (LOCA) provides a bounding analysis for this event. The limiting LOCA event for the MNGP as analyzed in accordance with 10CFR50 Appendix K conditions is based upon single failure of the Low Pressure Coolant Injection (LPCI) injection valve, effectively making LPCI inoperable for the event. The large break Design Basis Accident (DBA) with LPCI injection valve failure (which is analytically equivalent to the condition of both LPCI subsystems being inoperable) is the event analyzed for the current Licensing Basis Peak Cladding Temperature (PCT). This analysis bounds the event as a recirculation pump flow mismatch event is less limiting than the LOCA with LPCI injection valve failure analysis. Therefore, this recirculating loop flow mismatched event is less limiting than a previously analyzed event and ENS 49312 may be retracted as an unanalyzed event. The licensee has notified the NRC Resident Inspector. Notified R3DO (Skokowski). | Unanalyzed Condition |
| ENS 48809 | 6 March 2013 10:01:00 | Momentary Loss of Shutdown Cooling | At 0401 CST on 3/6/2013, while in RHR-High (Residual Heat Removal-High) water level the plant experienced a momentary Loss of Shutdown Cooling which resulted in a loss of safety function for Residual Heat Capability. Division 2 RHR shutdown cooling was restored within approximately 90 seconds without issue. No changes were experienced in refuel volume temperature or level during the loss of RHR shutdown cooling. This occurred shortly after a flow adjustment on the system was made utilizing the outboard valve. The inboard valve was reopened and an investigation is in progress. At the time of the valve closure, decay heat removal continued from Reactor Water Cleanup in heat reject mode and fuel pool cooling (with the fuel pool gates removed) is in service. Division 1 RHR (Shutdown Cooling) was available (not Operable) at the time of the loss. It is not currently understood why the injection valve closed. All systems functioned as required except for the spurious closing of MO-2015 (the Div 2 RHR inboard injection valve). The following make-up sources are available: Divisions 1 and 2 RHR, Divisions 1 and 2 Core Spray, CRD (Control Rod Drive), CST (Condensate Storage Tank) via a Core Spray with pressurizing station bypassed. The licensee notified the NRC Resident Inspector.
On March 6, 2013 (Notification No. 48809) NSPM (Northern States Power Monticello) reported in accordance with 10 CFR 50.72 (b)(3)(v)(B), a momentary closure of valve MO-2015 in the operating Residual Heat Removal (RHR) subsystem as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function. Following the event, the RHR SDC (shut down cooling) subsystem was removed from operation for equipment forensics and troubleshooting. Results validated that valve MO-2015 was operable and no issues were identified with the associated electrical circuitry, or the RHR SDC subsystem. The decay heat removal requirements of LCO 3.9.7, RHR - High Water Level, were met and there was not a loss of safety function. Therefore, NSPM retracts the March 6, 2013 notification for this event. The licensee notified the NRC Resident Inspector, state and local authorities, and may make a press release. Notified the R3DO (Passehl). | Time of Discovery |
| ENS 48616 | 20 December 2012 22:00:00 | Unanalyzed Condition Due to an Identified Degraded Fire Barrier | During a walkdown on December 20, 2012 at 1600 CST, two degraded Appendix R fire barriers (walls) were identified. These barriers separate the Torus Room (Fire Area IV)/ 'A' RHR Room (Fire Area I) and the Torus Room (Fire Area IV)/ 'B' RHR Room (Fire Area II). The walls separate Appendix R fire safe shutdown divisional equipment. A fire watch was established as a compensatory measure immediately following identification of the issue on December 20, 2012. The barrier affecting the 'B' RHR Room has been repaired on both sides. The barrier affecting the 'A' RHR Room has been repaired on the Torus Room side. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). The fire watch remains in place until verification of the completed repair is performed. The licensee notified the NRC Resident Inspector.
An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on December 21, 2012 for degraded fire barriers between the Torus Room (Fire Area IV) and 'A' RHR Room (Fire Area I), and the Torus Room (Fire Area IV) and 'B' RHR Room (Fire Area II). Subsequent engineering analysis determined that the degraded fire barriers 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 licensee will notify the NRC Resident Inspector. Notified R3DO (Kunowski). | Safe Shutdown Unanalyzed Condition Fire Barrier Fire Watch |
| ENS 48190 | 15 August 2012 01:45:00 | Safety System Overpressure Protection Failure Due to Closed Valves | At 2045 (CDT) on 8/14/12, MNGP (Monticello Nuclear Generating Plant) Operations determined that valves RHR-82 and RHR-84 had been inappropriately closed as part of an isolation clearance order for work on shutdown cooling suction piping. These valves are required to be open to provide overpressure protection for RHR piping passing through primary containment penetration X-12. Upon discovery of the condition, Primary Containment was declared Inoperable and the Required Actions of Tech Spec 3.6.1.1 were entered. Following discovery, the isolation was restored and the valves opened. At 0001 (CDT) on 8/15/12, Primary Containment was declared Operable. This issue is being reported in accordance with 10CFR50.72(b)(3)(v)(C) and 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety functions of a system needed to control the release of radioactive material or to mitigate the consequences of an accident. The MNGP Senior NRC Resident Inspector has been notified of this issue. The licensee will contact the Minnesota State Duty Officer.
This notification is a retraction of ENS 48190 based on further engineering evaluation. Monticello had previously evaluated penetration X-12 for thermally induced over pressurization. The evaluation qualified the piping components in the penetration for a maximum pressure of 3,306 psig using ASME Section III Appendix F operability criteria. The peak pressure calculated for the penetration was 2,743 psig based on Reactor pressure of 1000 psig with Reactor in Mode 1, and at worse case LOCA conditions for the Drywell. These assumptions and parameters envelop those that were present when valves RHR-82 and RHR-84 were closed on August 14, 2012. Therefore, this event would not have prevented the fulfillment of the safety function reported. The NRC Resident Inspector has been notified. Notified R3DO (Duncan). | Time of Discovery |
| ENS 48072 | 5 July 2012 17:58:00 | Unanalyzed Condition Due to Blocking Both Reactor Building Railroad Bay Doors | At 1258 CDT on 07/05/2012, Operations was notified that both panels of Door 45 (south doors for the reactor building railroad bay airlock) were blocked by a man lift. Blocking both doors represents an unanalyzed condition as a flow path through the door is assumed for pressure relief during postulated HELB events. The man lift was immediately removed correcting the situation and all work related to Door 45 was stopped. Door 45 was blocked for approximately 20 minutes. The NRC Resident Inspector has been notified.
The purpose of this notification is to retract the previous Event Notification Report (#48072) made by the Monticello Nuclear Generating Plant on 7/5/2012. The initial report indicated that blocking both panels of the railroad bay doors by a man lift represented an unanalyzed condition, as a flow path through the door is assumed for pressure relief during postulated HELB events, and was reported in accordance with 10CFR50.72(b)(3)(ii)(B), Unanalyzed Condition. Since the initial report, Engineering has completed an evaluation that demonstrates equipment supporting safe shutdown would have been capable of performing their specified design function during postulated HELB events. Based on this analysis, the condition initially reported in EN #48072 did not result in an unanalyzed condition that significantly degraded plant safety and is therefore retracted. The NRC Resident Inspector has been informed of this retraction. Notified R3DO (Cameron). | Safe Shutdown Unanalyzed Condition |
| ENS 46327 | 11 October 2010 18:05:00 | Technical Specification Does Not Account for Power Uprate | On October 11, 2010 at 1305 CDT it was identified that the analysis of record for the Technical Specification 3.3.5.1, Table 3.3.5.1-1 function 1e and 2e, Reactor Steam Dome Pressure Permissive-Bypass timer (Pump Permissive) did not reflect current plant conditions. Specifically, the analysis was not updated to account for any increase in plant licensed power and a change to the RWCU (Reactor Water Cleanup System) isolation for enhanced ability to isolate RWCU on a line break on critical crack. The allowable value for these function is greater than or equal to 18 minutes and less than or equal to 22 minutes. All equipment associated with emergency core cooling function are unaffected. Discussion with General Electric indicates that a margin exists to accommodate the higher power level. Additionally, the changes to the RWCU isolation logic added leak detection instruments that will isolate RWCU earlier for the majority of pipe leaks. This discovery is being reported as an unanalyzed condition solely due to the lack of a formal analysis of current plant conditions. The NRC Resident Inspector has been notified.
Under NUREG-0737, Item II.K.3.18 is a regulatory requirement to implement a modification to extend the ADS (Automatic Depressurization System) to a unique event sequence that involves multiple failures including HPCI (High Pressure Coolant Injection) plus no operator action after 10 minutes. According to Item II.K.3.18, the bypass timer logic complements, but does not replace, the existing ADS actuation logic. This requirement is not associated with any design basis accident mitigation sequence at MNGP (Monticello Nuclear Generating Plant). The plant has performed an evaluation that addresses the changes in plant thermal power and RWCU (Reactor Water Cleanup System) enhanced isolation capabilities to the analysis of record. This evaluation concluded that Peak Clad Temperature (PCT) will remain under the 2200 ?F acceptance limit with the current Technical Specification allowable value, current plant configuration and current licensed thermal power. GEH (General Electric) did an independent evaluation that concluded that based on use of limiting scenarios analyzed for MNGP, this condition is only a lack of formal analysis of current plant conditions and that no Substantial Safety Hazard exists. It is judged that the maximum Reactor Steam Dome Pressure Permissive-Bypass timer (Pump Permissive) setting of 22 minutes will not result in a predicted PCT higher than 2200?F with consideration of the RWCU pipe break isolation instrumentation modification. Since the timers with their current setpoint will protect the fuel cladding, this event does not significantly degrade safety. Therefore, the event notification made on October 11, 2010 is being retracted. The licensee has notified the NRC Resident Inspector. Notified R3DO (Skokowski) | Unanalyzed Condition Fuel cladding Power Uprate |
| ENS 45424 | 10 October 2009 18:21:00 | Control Room Filtration Out of Service for Maintainance | Post maintenance operability testing of the 'A' CREF (Control Room Emergency Filtration) subsystem will result in a planned potential loss of safety function for the CREF for a brief period of time when the 'B' CREF subsystem is simultaneously made inoperable during the testing. Clear guidance for timely restoration of the 'B' CREF subsystem, and therefore, CREF safety function is included in the test procedure. An operator will be dedicated to the testing to ensure that the CREF will perform its safety function if required. The licensee will notify the NRC Resident Inspector.
Monticello is retracting the event reported based on further evaluation. An investigation of the event found the removal of the (CREF) system from service was part of a planned evolution for maintenance or surveillance testing and done in accordance with an approved procedure and the plant's TS (Technical Specifications). In addition, the event would not have prevented the completion of the fulfillment of a safety function since an operator was stationed and briefed that in the event of the start of any transient, the CREF system would be immediately restored to operability and thereby ensure the train would have been available to perform its safety function In its required timeframe. The licensee has notified the NRC Resident Inspector. R3DO (Riemer) notified. | |
| ENS 44957 | 2 April 2009 10:43:00 | Residual Heat Removal Inoperable | At 0543 (CDT) on April 2, 2009 at the Monticello Nuclear Generating Plant (MNGP) an Operator made the following discovery during performance of his rounds. The flowrate and pressure of the #14 Residual Heat Removal Service Water (RHRSW) pump motor cooling appeared to be low. Investigation found the flow to be approximately 1 gpm. The cooling water supply flow to the pump motor cooler comes from either RHRSW or Service Water. Based on these indications, Operators declared the RHR Shutdown Cooling inoperable and entered actions for Technical Specifications 3.9.7. Actions have been completed to provide an alternate water supply to the RHRSW pump motor oil cooler. Based on the inoperability of the RHR Shutdown Cooling, this event is reportable under 10 CFR 50.72(b)(3)(v), 'An Event or Condition that could have Prevented the Fulfillment of a Safety Function-Capability to Remove Residual Heat.' The station is performing further investigation into the event and will develop corrective actions based on the results of the investigation. The Licensee notified the NRC Resident Inspector.
Monticello is retracting the event reported based on further evaluation. An investigation of the event found test data that demonstrates the RHRSW pump would not have lost its ability to provide cooling water to the shutdown cooling system and therefore no loss of safety function occurred. The test data provides documentation that the thrust bearing oil bath temperature of the RHRSW pump motor would not have exceeded the 200 deg F limit imposed by the motor supplier (GE) at the flow rate found by the operator. The test data indicated with the cooling water at a flow rate of less than 0.9 GPM, at 65 deg F the service water temperature and flow would be sufficient to maintain the motor oil bath temperature below 200 deg F. During the event, the actual event parameters (cooling water flow rate >1 gpm and temperatures< 65 deg F) were less severe than the test parameters and therefore are bounded by the test. Since there was no impact on the RHRSW system's ability to provide cooling water to the RHR system, the RHR system maintained the ability to provide shutdown cooling and residual heat removal. Therefore the event can be retracted since the condition that was reported in the initial event notification report would not have resulted in the prevention of the fulfillment of a safety function (residual heat removal). The licensee notified the NRC Resident Inspector and will notify the State of Minnesota. Notified R3DO(Lara) | |
| ENS 44723 | 15 December 2008 22:16:00 | Unanalyzed Condition Due to Steam Chase Temperature Greater than 165F | At 1616 on 12/15/08, a plant heating boiler trip resulted in a loss of a reactor building ventilation. The loss of reactor building ventilation resulted in maximum average main steam chase temperatures greater than or equal to 165F. High energy line break (HELB) analysis of piping in the steam chase assumes an initial average temperature prior to the break of 165F. Temperature greater than or equal to 165F in the steam chase challenges EQ qualification of the piping analysis. Abnormal procedures for loss heating boiler and ventilation system failure were entered. C.3 (Shutdown) and C.5-1300 (secondary containment control) were also entered. The plant heating boiler was restarted and ventilation restored prior to power reduction. All systems have been returned to normal. The licensee notified the NRC Resident Inspector.
The licensee is retracting this report based on the following: Monticello is retracting the event reported based on further evaluation, which found that the issue was not an unanalyzed condition that seriously degraded plant safety. The investigation of the event found the peak temperature achieved was 167.2 degrees F and the condition lasted for approximately 11 minutes. Engineering review of Safety System Components found no impact on the equipment for the temperature reached, Additionally, revised High Energy Line Break (HELB) calculations performed with an initial average Steam Chase Room temperature of 180 degrees F before a HELB determined that Safety System components could perform their safety functions. The station has identified the cause for the event and corrective actions will be tracked in the station's corrective action program. Since there was no impact on the equipment in either Environmental Qualification (EQ) or safety function, the temperature of the event was less than the revised calculation temperature, and the unanalyzed condition that existed in the initial event notification report no longer exists and did not result in a condition that seriously degraded plant safety, this event can be retracted. The licensee informed the NRC Resident Inspector. Notified R3DO (Ring). | Unanalyzed Condition |
| ENS 42013 | 25 September 2005 19:10:00 | Degradation in Fire Barrier Discovered During an Inspection | During an inspection of structural steel in the Emergency Filtration Train (EFT) building, it was determined that a portion of the steel did not have adequate fire retardant material protecting the steel. This condition would have compromised the 3 hour-fire barrier between the 2 divisions of the Emergency Service Water (ESW) System, impacting # 11 Emergency Diesel Generator ESW Pumps, # 13 and # 14 ESW Pumps. This condition could have resulted in the inability to establish and maintain cold shutdown conditions in the event of a fire in this area. A fire impairment was entered and an hourly fire watch was established to address this condition. All ESW divisions are presently operable. The NRC Resident Inspector will be notified of this event by the licensee.
Monticello is retracting the event notification based on further investigation of the issue. The station has completed calculations that confirm the affected steel beams are not required to maintain fire barrier integrity. Further, the fire severity in the zone will not cause the steel beams to fail. Therefore, the ability of the station to establish and maintain cold shutdown conditions in the event of a fire in this area was not impacted. Based on this information, Monticello has determined there was no unanalyzed condition as reported in Event Notification # 42013. The degraded fire retardant material has been replaced. The licensee notified the NRC Resident Inspector. | Unanalyzed Condition Fire Barrier Fire impairment Hourly Fire Watch |
| ENS 40481 | 26 January 2004 20:30:00 | Recirculation Fan Alteration Affects Accident Mitigation | V-ERF-11 ('A' Emergency Filtration Train (EFT) recirculation fan) was found to have an improper alteration affecting the fan's shaft speed, and 'A' EFT was declared inoperable. Concurrently, the #12 Emergency Diesel Generator (EDG) was inoperable for planned maintenance, making 'B' EFT inoperable. This condition is a loss of safety function during a design basis accident, and impacts the ability of the plant to mitigate the consequences of an accident. The licensee notified the NRC Resident Inspector.
At 1958 CST the licensee declared the #12 EDG operable and thus the #12 ("B") EFT was also operable. Additionally, at 2010 the licensee declared the #11 ("A") EFT operable. Notified R3DO (Burgess).
Based on further investigation V-ERF-11 ('A' Emergency Filtration Train (EFT) recirculation fan) would have been able to provide the required flow and would have fulfilled its required safety functions with the improper alterations. Therefore the event was not reportable. Notified R3DO ( O'Brien). | |
| ENS 40146 | 10 September 2003 22:46:00 | Unanalyzed Condition Due to a Potential Break in Fire Protection Water System | Potential break of Fire Water Main in Admin Bldg could cause both divisions of Safe Shutdown equipment to be inoperable by flooding the battery rooms for #11 & 12 125VDC Batteries. Vulnerable section of piping has been isolated to eliminate the flooding concern. Fire Protection compensatory actions have been established in accordance with the site Fire Protection Program. The licensee informed the NRC resident inspector.
Monticello is retracting the event reported based on evaluations which indicate that the fire main is not considered a potential flooding source. Additional evaluations are ongoing, and issues will be entered into the station's corrective action program. The licensee has notified the NRC Resident Inspector. Notified R3DO (Hills) | Safe Shutdown Unanalyzed Condition Fire Protection Program |
| ENS 40008 | 22 July 2003 14:48:00 | Helb Door Not Latched as Required Due to Personnel Error | A High Energy Line Break (HELB) door was not latched as required. This condition is being reported as an event that could have prevented the fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(A). The door is currently closed. The HELB door which separates two critical Motor Control Center (MCC) areas was unlatched for less than two (2) minutes. The licensee will inform the state representative and has informed the NRC resident inspector.
Because plant safety was not significantly degraded, this event is not reportable under the unanalyzed condition criteria based on: (1) the door in either event was in an uncontrolled condition for less than one minute, (2) the door was not materially affected, only operated improperly, (3) the PRA significance of the event was low, and (4) the HELB Barrier door was not open for a period longer than is allowed by station procedural guidance. R3DO (C. Miller) notified. The station continues to review the event in the station's corrective action program. The NRC Resident Inspector was notified of this retraction by the licensee. | Unanalyzed Condition |