Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55896 | 13 May 2022 16:11:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Control Room Envelope Inoperable | The following information was provided by the licensee via email: On 5/13/22 at 1111 CDT the station entered LCO 3.7.4 Condition B for Control Room Envelope being inoperable. This was due to results from an inspection in the Steam Jet Air Ejector room that identified steam leakage exceeding the leakage rate assumptions made in the Alternate Source Term (AST) dose analysis calculation. Therefore, this is being reported in accordance with 10CFR50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades plant safety and 10CFR50.72(b)(3)(v)(D) for any event or condition that at the time of discovery could have prevented fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. There is no impact to the health and safety of the public. NRC Resident has been notified. | Steam Jet Air Ejector Control Room Envelope | |
| ENS 54111 | 11 June 2019 16:32:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Normally Closed Primary Containment Isolation Valves Found in the Open Position | At 1132 CDT on 6/11/2019, both manual primary containment isolation valves in a one-inch service air line were found open. This resulted in an open primary containment penetration. Both valves are required to be closed for Primary Containment Isolation Valve Operability. Both valves were closed and independently verified closed at 1149 CDT on 6/11/2019. This is being reported under 10 CFR 50.72(b)(3)(v)(C) and (D), and 10 CFR 50.72(b)(3)(ii)(B). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee also notified the State of Minnesota State Duty Officer. | Primary containment | |
| ENS 53997 | 12 April 2019 23:15:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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). | Core Spray Low Pressure Coolant Injection | |
| ENS 53597 | 11 September 2018 05:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Unanalyzed Condition | On 9/10/2018, the 11 Core Spray (CSP) loop was placed in service to support quarterly surveillance testing. With the 11 CSP pump in service it was identified that the check valves isolating the 11 CSP system from the keep fill supply were leaking by. At 1129 CDT on 9/11/2018, it was identified that this leakage may have exceeded the leakage rate assumptions made in the dose analysis calculation for emergency core cooling system (ECCS) leakage outside containment following a loss of coolant accident (LOCA). Therefore, this is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The potential ECCS leak pathway has been isolated. There is no impact to health and safety of the public. The NRC Resident Inspector has been notified. | Core Spray Emergency Core Cooling System | |
| ENS 52814 | 20 June 2017 04:53:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection Inoperable | At 2353 CDT on 6/19/2017, while performing the High Pressure Coolant Injection (HPCI) quarterly surveillance following planned 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 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. Investigation into the failure of the HPCI system to start is in progress. The unit remains at 100% power. The health and safety of the public was not affected. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection | 05000263/LER-2017-004 |
| ENS 52454 | 21 December 2016 15:35:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 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). | High Pressure Coolant Injection | |
| ENS 52396 | 27 November 2016 20:47:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection Declared Inoperable | At 1447 (CST) on 11/27/2016 while troubleshooting a minor leak on the High Pressure Coolant Injection (HPCI) turbine, it was discovered that the HPCI turbine exhaust drain pot high level bypass switch was not functioning per design to support removal of condensate from the HPCI turbine casing. This resulted in some water accumulation within the HPCI turbine casing. Subsequently, HPCI was declared INOPERABLE and 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. The plant remains at 100 percent power with no challenges to the health and safety of the public. The NRC Resident Inspector has been notified. Technical Specification limiting condition for operation requires HPCI to be Operable within 14 days. The licensee will be notifying the State of Minnesota regarding the event. | High Pressure Coolant Injection | 05000263/LER-2016-003 |
| ENS 51812 | 22 March 2016 06:04:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection (Hpci) Oil Leak | On 3/22/2016 during performance of HPCI FLOW CONTROL SYSTEM DYNAMIC TEST PROCEDURE, an oil leak was discovered on the hydraulic control oil piping. HPCI had previously been declared INOPERABLE due to planned maintenance, however as a result of the oil leak HPCI remains INOPERABLE. This oil leak would have cause HPCI to be declared INOPERABLE had it been found outside of the planned maintenance. The plant remains at 100% power with no challenges to the health and safety of the public. HPCI is in a 14 day technical specification to repair the oil leak. The licensee notified the NRC Resident Inspector. | High Pressure Coolant Injection | 05000263/LER-2016-001 |
| ENS 50915 | 21 March 2015 10:37:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 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). | High Pressure Coolant Injection | |
| ENS 50899 | 16 March 2015 23:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 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). | High Pressure Coolant Injection | |
| ENS 50705 | 29 December 2014 02:23:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Two Inoperable Emergency Diesel Generators | While the 12 Emergency Diesel Generator (EDG) was inoperable for performance of the monthly surveillance, adjustments were inadvertently made to 11 EDG which made it inoperable. As a result, Technical Specification (TS) 3.8.1 Condition E, for both EDG's inoperable was entered. Monticello has subsequently restored 12 EDG to an operable status within the 2 hour TS LCO (Limiting Condition for Operation) completion timer requirement. The station remained in a safe condition during this discovery with 12 EDG available at all times. The plant continues to operate in a normal condition with no initiating events present. The health and safety of the public was not impacted as a result of this condition. The NRC Resident Inspector has been notified. EDG 12 was restored to operable status at 2214 CST and EDG 11 will remain inoperable until a surveillance test is performed to start the EDG and restore the local governor control idle speed to the correct setting. The licensee will be notifying the Minnesota State Duty Officer. | Emergency Diesel Generator | |
| ENS 50496 | 26 September 2014 03:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Containment Isolation Declared Inoperable Due to Relay Age | At 2200 CDT on September 25, 2014, the Duty Shift Manager was notified that Agastat relays associated with Primary Containment Isolation valves on the Hydrogen-Oxygen Analyzing System are beyond the analyzed shelf life for relays that are in the normally energized state and are considered INOPERABLE. This affected both primary containment isolation valves for a containment penetration on multiple flow paths. This issue was determined to be reportable under (10 CFR) 50.72 (b)(3)(v)(C) & (D) for an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material and mitigate the consequences of an accident. Additionally, the required actions involved isolating six flow paths via manual isolation valves. This action rendered the Hydrogen-Oxygen Analyzers non-functional for both trains and constitutes a loss of Emergency Preparedness and Accident Assessment Capability. This is reportable under (10 CFR) 50. 72(b)(3)(xiii). The Primary Containment Isolation Valves have been, and remain, in their closed position to satisfy their Primary Containment Function and protect the health and safety of the public. The NRC Senior Resident Inspector has been notified. The licensee will notify the State of Minnesota. The relays of concern were manufactured 19 years ago and have been in operation for 11 years, versus a manufacturer assumption of a 10 year operational lifespan.
Partial retraction for EN 50496. This is an update of Emergency Notification System (ENS) report 50496 that was submitted at 0253 EDT on Friday, September 26, 2014. ENS notification was made due to four relays associated with the sampling valves on the Hydrogen-Oxygen Analyzing (HOA) system that perform Primary Containment Isolation Valve (PCIV) functions. These relays were discovered installed beyond their manufacturer qualified service life, which called operability into question. The portions 10 CFR 50.72 (b)(3)(v)(C) & (D) are being retracted after subsequent bench testing and investigation of system operability. Based on the past operability evaluation, all four relays associated with PCIV functions on the HOA system would have performed their specific safety function of primary containment isolation, as required by the facility's technical specifications. Therefore, this event does not meet the threshold of an event or condition that would prevent fulfilment of a safety function. The loss of emergency preparedness and accident assessment capability previously reported under 10 CFR 50.72 (b)(3)(xiii) remains unchanged. The NRC Resident Inspector has been notified. Notified R3DO (Peterson). | Primary containment | |
| ENS 50345 | 5 August 2014 19:46:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Trains of Control Room Emergency Filtration System Inoperable | The Division 1 Control Room Emergency Filtration System (CREF) was inoperable for scheduled replacement of charcoal. During the scheduled maintenance, Division 2 CREF was placed into service. Approximately 5 minutes after startup (1446 CDT on 8/5/2014), the Division 2 CREF recirculation fan tripped off for unknown reasons. This rendered both trains of CREF inoperable. This required entry into Technical Specification TS 3.0.3. This is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (D) Mitigate the consequences of an accident. At 1707 CDT on 8/5/2014, the Division 1 CREF train maintenance was completed and the Division 1 CREF was declared operable. TS 3.0.3 was exited at this time. Investigation is in progress to determine the cause of the Division 2 CREF trip. The control room boundary was not challenged during this time period with any change in radiation levels as plant operation was unaffected. Thus, the health and safety of the public was not affected. The licensee notified the NRC Resident Inspector and the State of Minnesota Duty Officer. | Control Room Emergency Filtration System | |
| ENS 49970 | 28 March 2014 18:58:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Failure of Secondary Containment Door Interlock | At 1358 (CDT) on March 28, 2014, the Control Room was notified that two Secondary Containment doors (DOOR-62 and DOOR-63) were open at the same lime. This occurred while two employees were entering and exiting the Reactor Building at the exact same time. The time that both doors were open was approximately one (1) second. Secondary Containment differential pressure was maintained throughout the event. With both doors open, technical specification surveillance requirement SR 3.6.4.1.3 was not met and Secondary Containment was declared inoperable. Secondary Containment was declared operable after independently verifying at least one Secondary Containment access door was closed. The health and safety of the public was maintained as the plant was in a normal condition with no initiating event in progress or signs or elevated radiation levels within Secondary Containment. The NRC Resident Inspector has been notified. | Secondary containment | 05000263/LER-2014-006 |
| ENS 49819 | 11 February 2014 20:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Drywell to Torus Vacuum Breaker Failure During Surveillance Testing | While cycling AO-2382A (TORUS-DW VAC BREAKER) for required surveillance testing, (the vacuum breaker) did not indicate fully closed on all available indicators. The procedure for this condition was utilized to continue to cycle the vacuum breaker to achieve closed indication on all available indicators. The vacuum breaker was cycled a total of four (4) times and dual indications were present for approximately six (6) minutes. During the six (6) minutes that the vacuum breaker indications did not show fully closed, the Technical Specification Limiting Condition for Operation (TS LCO) requirement was not met. The Monticello Safety Analysis assumes all eight (B) vacuum breakers are closed, therefore this condition is being reported per 10CFR50.72(b)(3)(ii)(B) and per 10CFR50.72(b)(3)(v)(D). The vacuum breakers are all capable of performing their design function and all safety related equipment is operable." The NRC Resident Inspector has been notified. The same vacuum breaker failed its surveillance test in a similar fashion on February 7, 2014 (See EN #49808). The surveillance test for this vacuum breaker is due again on February 18, 2014. | 05000263/LER-2014-003 | |
| ENS 49808 | 7 February 2014 16:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Single Drywell to Torus Vacuum Breaker Not Going Fully Closed During Surveillance Test | After cycling AO-2382A (Drywell to Torus Vacuum Breaker) for surveillance testing, it did not indicate fully closed. The procedure for this condition was entered and after cycling the valve several times, the vacuum breaker indicated full closed. During the approximately eight minutes that the indication showed that it was not closed, the Technical Specification Limiting Condition for Operation (LCO) requirement was not met. After validation that the vacuum breaker had opened as required, and was closed successfully, the safety function was restored. The health and safety of the public was not jeopardized as the plant was in a normal condition and an initiating event was not in progress. The USAR (Updated Safety Analysis Report) assumes all eight vacuum breakers to be closed. This condition therefore put the nuclear power plant in an unanalyzed condition and is reportable per 10CFR50.72(b)(3)(ii)(B). This condition, at time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident and is reportable per 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector, the State of Minnesota Duty Officer, and the local counties. | 05000263/LER-2014-002 | |
| ENS 49363 | 19 September 2013 22:50:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 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). | Service water High Pressure Coolant Injection | |
| ENS 49356 | 19 September 2013 03:29:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Both Secondary Containment Access Doors Briefly Opened Simultaneously | While performing the secondary containment airlock door interlock surveillance, the interlock to the main plenum room did not prevent the opening of both doors to the plenum room airlock (DOOR-85 and DOOR-86). The plenum room airlock doors were immediately closed. The time both doors were opened is estimated to be approximately one (1) second. When both doors open, Technical Specification surveillance requirement SR 3.6.4.1.3 was not met and secondary containment was declared inoperable. Secondary containment was declared operable after independently verifying at least one secondary containment access door was closed. There were no radiological releases associated with this event. The NRC Resident Inspector has been notified.
This update provides additional information on the initial notification of the event. On 9/18/13, while testing secondary containment airlock doors, the interlocks did not prevent opening of both doors simultaneously. With the outer door to the main plenum room open, the inner door was able to be opened. At this point, Technical Specification SR 3.6.4.1.3 was not met and secondary containment was inoperable. The inner door was closed immediately. While in this condition, the inner door was then opened, and the interlock did not prevent the opening of the outer door. The outer door was closed immediately. Secondary containment was declared operable after verifying at least one of the airlock doors was closed. There were no radiological releases associated with this event. The NRC Resident Inspector has been notified. Notified R3DO (Reimer). | Secondary containment | |
| ENS 49113 | 13 June 2013 19:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Emergency Diesel Generators Start | While preparing for an equipment test Thursday afternoon, Monticello Nuclear Generating Plant lost off-site power on its normal off-site power feed. Power for safety related loads was automatically transferred to the alternate off-site power source. The Emergency Diesel Generators started as designed but did not load onto the safety related busses due to the availability of off-site power. Operators stabilized the plant, which is shutdown for a refueling and maintenance outage, in less than an hour and are investigating the cause of the event. The current plant focus is on restoring the normal off-site power feed. The event posed no danger to the public or plant workers, and no one was injured. There was no release of radiation. Plant safety systems continue to be powered by the backup off-site power feed, with the emergency diesel generators available if needed. Event Specifics: At approximately 1430 CDT, during a refueling outage with the plant in Mode 4, reactor level at approximately 200 inches, and a full Scram already inserted, a loss of normal off-site power occurred due to a fault in a non-safety related bus supply breaker. The fault was in the 13.8 KV supply breaker to the #11 bus. This caused the Station 2R transformer to lockout, resulting in a loss of the normal off-site power to Essential Busses 15 and 16. Shutdown Cooling (SDC) was lost for approximately 1 hour due to loss of supply power and isolation of the common suction valves. Both 11 and 12 Emergency Diesel Generators (EDGs) automatically started but did not load onto their respective busses (as designed) due to the 1AR emergency off-site transformer re-energizing both 15 and 16 bus. This essential bus transfer is being reported as a 'Valid actuation of emergency AC electrical power systems' under 10CFR50.72(b)(3)(iv). During the event the decision was made to shut down the EDGs which rendered them inoperable for a short period of time until the Fast Start capability was reset. The period of time that the EDGs were inoperable is being reported as a 'Condition that could have prevented the fulfillment of the safety functions to remove residual heat, control the release of radioactive material, and mitigate the consequences of an accident under 10CFR50.72(b)(3)(v)(B), (C), and (D). Both EDGs have been restored to Automatic Standby Status and are operable. The loss of power resulted in a Group II Containment Isolation signal causing secondary containment to isolate and Standby Gas Treatment and Control Room Emergency Filtration to initiate as well as associated Group II Containment Isolation Valves to close. This is being reported as a 'General containment isolation signal ESF actuation' under 10CFR50.72(b)(3)(iv). The containment isolation has been reset, and SDC and SFPC have been restored. Reactor temperature rose approximately 4 degrees F during the event from 161 degrees to 165 degrees which remained in the prescribed operating band. Reactor level did not change. The licensee has notified the NRC Resident Inspector. | Secondary containment Emergency Diesel Generator Shutdown Cooling | |
| ENS 48190 | 15 August 2012 01:45:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | 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). | Primary containment Shutdown Cooling | |
| ENS 47479 | 27 November 2011 22:57:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Rod Worth Minimizer Control Switch Found Out of Required Position | After transitioning to Mode 2 from Mode 4, while performing the Rod Worth Minimizer (RWM) operability test, it was discovered that the RWM control switch was in the BYPASS position. The RWM enforces predetermined control rod withdrawal and insertion sequences. Complying with these predetermined sequences ensures a Control Rod Drop Accident does not exceed analytical limits. With the control switch in the BYPASS position, the RWM was inoperable and would not have enforced the predetermined control rod withdrawal sequence. The RWM control switch was restored to the OPERATE position and the RWM was verified to be operable. This issue is being reported under 50.72(b)(3)(v)(D) as an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of the RWM, which is a system needed to mitigate the consequences of the Control Rod Drop Accident. The licensee will be notifying the NRC Resident Inspector. | Rod Worth Minimizer Control Rod | |
| ENS 47306 | 29 September 2011 22:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Inadequate Surveillance Testing of Emergency Diesel Generators | Monticello has discovered that it has not met Technical Specification Surveillance Requirement (SR) 3.8.1.7 relating to the largest single post-accident load reject for the Emergency Diesel Generators (EDG). Although the current test designated post-accident load is successfully load rejected during the surveillance, the test load rejection must be higher to bound all post-accident load scenarios. The capability of an EDG subsystem to recover from a reject of the largest single post-accident load testing has not met the requirements of SR 3.8.1.7. Therefore, both EDGs have been declared inoperable. Both EDGs are considered Functional and Available for use at this time. There were no automatic EDG initiation signals associated with this event. The licensee notified the NRC Resident Inspector and will notify the State. | Emergency Diesel Generator | |
| ENS 46999 | 30 June 2011 10:16:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Power Range Monitor Channels Out of Alignment | At 0516 (CDT) on 6/30/11 after control rod movements to support rod pattern adjustment, 3 of 4 APRMs were out of required Technical Specification deviations of +/-2% power in relation to calculated Core Thermal Power. APRM #1 was at -3.6% deviation, APRM #3 was at +2.5% and APRM #4 was at +3.1 %. APRMs 1, 3, and 4 were declared inoperable. With 3 of the 4 APRM channels affected, the functions of the APRM were inoperable and that RPS trip capability had not been maintained. Technical Specification Conditions 3.3.1.1.A and 3.3.1.1.C were entered at 0516. All three (3) APRM gains were adjusted and the Tech Spec Conditions were exited at 0540 (CDT). Thermal Limits were evaluated and no limits were challenged. This event is reportable under 10CFR50.72 (b)(3)(v) as an event that could have prevented the fulfillment of the safety function of a system needed to: 50.72(b)(3)(A) shutdown the reactor and 50.72(b)(3)(D) mitigate the consequences of an accident. The licensee notified the NRC Resident Inspector. | Control Rod | |
| ENS 46937 | 8 June 2011 13:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Secondary Containment Inoperable Due to Ventilation Alignment Issue | Secondary containment was declared inoperable after transferring refuel floor supply fans. Secondary containment D/P (Differential Pressure) lowered to 0.17 inches of water vacuum which does not meet the surveillance requirement to have secondary containment vacuum greater than or equal to 0.25 inches of water vacuum. Refuel floor ventilation was restored back to the previous configuration and secondary containment D/P was restored back to greater than 0.25 inches of water vacuum. Vacuum was less than 0.25 inches of water for approximately 4 minutes. There were no actual radiological releases associated with the event. Actual secondary containment integrity was not challenged. The lowered secondary containment D/P was a result of a ventilation lineup change. The licensee has notified the NRC Resident Inspector and the State of Minnesota. | Secondary containment | |
| ENS 46606 | 11 February 2011 09:27:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Inoperable Containment Isolation Dampers | On February 11, 2011 at 0327 (CST), Secondary Containment isolation damper V-D-61 (Reactor Building Outboard Isolation Damper) was discovered iced closed with the actuator broken. The corresponding inboard damper, V-D-62, was found blocked partially open due to icing. Technical Specification Condition 3.6.4.2.B was entered for one Secondary Containment penetration flow path with two isolation valves inoperable. This resulted in a condition which could have prevented the fulfillment of a safety function required to control the release of radioactive material and mitigate the consequences of an accident. The required action to isolate the flow path by use of one closed and de-activated valve was completed at 0354 (CST), within the four hour completion time. Repair activities are in progress. The licensee notified the NRC Resident Inspector and the State of Minnesota Duty Officer. | Secondary containment | |
| ENS 46431 | 22 November 2010 23:47:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Valves Identified Open in Mode 2 That Should Have Been Closed | At 1547 CST, November 22, 2010, the reactor mode was changed from Mode 4 to Mode 2 with the main steam drain valves, which are primary containment isolation valves, tagged open with power removed from their respective breakers. The valves were tagged to comply with S/D (Shutdown) operating procedure requirements. The valves should have been restored prior to making the mode change. The main steam line drain isolation valves MO-2373 and MO-2374 were restored at 1747 CST and verified at 1755 CST. Startup was held for determination of further actions needed. Reactor startup recommenced at 2200 CST. Currently the reactor is not critical. The licensee notified the NRC Resident Inspector. The licensee notified the Minnesota State Duty Officer. | Primary containment Main Steam Line Main Steam | |
| ENS 46397 | 5 November 2010 04:25:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Four Average Power Range Monitors Outside Allowable Range | Following a power reduction to 49 percent reactor thermal power at approximately 2312 (CDT) on November 4, 2010, 'B' high pressure feedwater heaters were isolated at approximately 2320 in preparation for repair of the 15B feedwater heater. Calculated core thermal power (CTP) rose by approximately 50 MWth. This resulted in all four average power range monitors (APRMs) failing to satisfy Technical Specification (TS) Surveillance Requirement SR 3.3.1.1.2 in that the absolute difference between the APRMs and the calculated power from the heat balance was greater than 2.0 percent rated thermal power (RTP). All four APRM channels were between approximately 3.5 and 4.0 percent lower than CTP. Since all four APRM channels were affected, the functions of the APRMs were inoperable and RPS trip capability had not been maintained. TS Conditions 3.3.1.1.A and 3.3.1.1.C were entered at 2325. All four APRM gains were adjusted and the TS Conditions were exited at 2349. This event is reportable under 10CFR50.72(b)(3)(v) as an event that could have prevented the fulfillment of the safety function of a system needed to: (A) shutdown the reactor and (D) mitigate the consequences of an accident. This event notification is being submitted outside the 8 hour reporting requirement. The licensee notified the NRC Resident Inspector.
The following is an update to Event Notification 46397 made on 5 November 2010 concerning Average Power Range Monitors (APRMs) inoperability due to >2% deviation from calculated Core Thermal Power (CTP). During review of inputs to the CTP calculation, Operations staff determined that indicated power levels were not consistent with the indicated electrical output of the turbine generator. Subsequent investigation revealed an erroneous feedwater flow input from the isolated 'B' feedwater heater string. Actual CTP was less than indicated CTP by approximately 50 MWth. This was validated and verified through manual calculation. 50 MWth constitutes approximately 3% of rated thermal power. With the adjustments made to APRM on 4 November 2010, this resulted in all four APRMs being outside of the acceptable tech spec value. Following Operations review, at 1108 on November 7, 2010, all four APRMs were declared inoperable and tech. spec. 3.3.1.1.A and 3.3.1.1.C actions were initiated. Values consistent with the actual plant configuration ('B' feedwater heater string isolated) were input to the CTP calculation with subsequent indicated CTP returning to actual CTP. All four APRM gains were adjusted and APRMs returned to operable status. Tech Spec 3,3.1.1.A and 3.3.1.1.C conditions were exited at 1207 on November 7, 2010. The licensee has notified the NRC Resident Inspector. Notified R3DO (Giessner). | Feedwater | |
| ENS 46394 | 4 November 2010 05:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Containment Air Lock Doors Not Operated Properly | On November 4, 2010, at 11:28 AM both doors in airlock 124 from secondary containment to access control were simultaneously open for a period <5 seconds. The doors were immediately closed. This condition resulted in an unplanned entry into Technical Specification 3.6.4.1.A for secondary containment. The condition could have prevented the Standby Gas Treatment System from developing a negative pressure within secondary containment following a design basis accident. This negative pressure is required to prevent ground consequences following an accident. The Standby Gas Treatment System remained operable throughout the event. The licensee was decreasing power at the time of the report for a condition unrelated to the report. The licensee will notify the Minnesota Duty Officer. The licensee notified the NRC Resident Inspector. | Secondary containment Standby Gas Treatment System | |
| ENS 46155 | 5 August 2010 16:45:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Momentary Loss of Secondary Containment Due to Personnel Passing Through Open Airlocks | On August 5, 2010 at 1145 CDT, both doors in Airlock 413 from Secondary Containment (SCT) to the 985 ft Radwaste Pump Room were simultaneously open for a period of approximately five (5) seconds and subsequently reclosed. This condition caused an unplanned entry into Technical Specification 3.6.4.1.A for SCT. The condition could have prevented the Standby Gas Treatment system from developing a negative pressure with SCT following a design basis accident. This negative pressure is required to prevent ground level release of radioactivity and to minimize onsite and offsite dose consequences following an accident. The Standby Gas Treatment system remained operable throughout the event. The licensee will inform the State and has notified the NRC Resident Inspector. | Secondary containment Standby Gas Treatment System | |
| ENS 45975 | 3 June 2010 15:50:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Secondary Containment Briefly Degraded | On June 3rd, 2010, at 1050 CST, both doors in Airlock 413 from Secondary Containment (SCT) to the Rad Waste 985' Pump Room were opened simultaneously for approximately five seconds and subsequently re-closed. This condition caused an unplanned entry into Technical Specification 3.6.4.1.A for SCT. The condition could have prevented the Standby Gas Treatment system from developing a negative pressure within SCT following a design basis accident. This negative pressure is required to prevent ground level releases of radioactivity and minimize onsite and offsite dose consequences following an accident, The Standby Gas Treatment system remained operable throughout the event. The site continues to assess the situation. The licensee has notified the NRC Resident Inspector and will also notify State authorities. | Secondary containment Standby Gas Treatment System | |
| ENS 45424 | 10 October 2009 18:21:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 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 45279 | 20 August 2009 13:38:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Both Divisions of Vital Switchgear Inoperable Because Flood Door Found Closed | At 08:38 on 08/20/09, a plant operator identified that DOOR-18, which is a normally open fire door, had closed due to a failed fusible link. With this door closed, the pathway for a potential flood due to a high energy line break (HELB) is blocked therefore closing off a drain path for the water. This represented an unanalyzed condition where both divisions of essential switchgear could be impacted. As a result, both divisions of essential switchgear were declared inoperable and Technical Specification LCO 3.0.3 was entered. With both switchgear divisions being inoperable, this condition is also an event that could have prevented fulfillment of a Safely Function and reportable under 50.72(b)(3)(v). At 0942 on 08/20/09, the closed fire door was restored to the open state. Both divisions of switchgear were declared Operable and LCO 3.0.3 was exited. No system actuations occurred as a part of this event. A continuous fire watch was posted as a compensatory measure while the fire door is being repaired. The licensee notified the NRC Resident Inspector. | ||
| ENS 45060 | 13 May 2009 02:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Instrument Valve Malfunction Potentially Affecting the Isolation Capability of One of the Main Steam Lines | The equalizing valve for one of the four Main Steam Line (MSL) Flow - High differential pressure switches on the 'B' MSL was leaking through. The leak effectively reduced the differential pressure across all four MSL Flow - High valves on the 'B' MSL. This reduction in differential pressure thus potentially would not allow the switches to isolate the 'B' MSL at the required setpoint. This switch was for group 1 isolation. This loss of safety function was restored by isolating the faulty valve block for DPIS-2-117A. All other switches are now reading normally. The repairs for the faulty valve are in progress. The licensee will notify NRC Resident Inspector, State, and local authorities. | Main Steam Line | |
| ENS 44050 | 10 March 2008 23:13:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Inoperable | During performance of the quarterly high pressure coolant injection (HPCI) Surveillance Test, a technical specification step could not be completed due to observed system water flow and discharge pressure oscillations. These oscillations are presently under investigation. The tech spec step involved establishing flow conditions at a certain discharge pressure. The problem is either with the test return system or the control system. A formal troubleshooting plan is being developed to determine the root cause and corrective action required to re-establish operability of the HPCI system. The system remains inoperable due to the problem found during testing. If the problem is found to be caused by the control system, then it could have potentially impacted the ability of the HPCI system to mitigate the consequences of an accident. HPCI is currently in a 14 day Tech Spec 3.5.1.h LCO. The licensee notified the NRC Resident Inspector. The licensee will be notifying the Minnesota Duty Officer.
Monticello is retracting the event reported based on further reviews of the event which found that the issue did not impact HPCI operability. A problem was identified with the test return valve CV-3503 which is not a safety-related component. The stations formal troubleshooting team has identified the cause for the degradation and corrective actions will be tracked in the station's corrective action program. The licensee will notify the NRC Resident Inspector. Notified R3DO (Cameron). | High Pressure Coolant Injection | |
| ENS 42607 | 27 May 2006 22:14:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Excess Temperature in Room Housing Safety Function Systems | At 1714 the Division II electrical buses were declared inoperable due to room temperatures greater than 104 F. The appropriate 24 hour LCO was entered. The C.4 abnormal procedure for loss of ventilation was entered. The procedure stated to consider opening doors in the area to provide additional ventilation. Doors were opened for additional cooling. The procedure states to declare ALL Division I 4KV equipment inoperable. At that time both 15 and 16 emergency buses were inoperable. At this time, this condition could have been prevented the fulfillment for Safety Function Systems needed to remove residual heat and mitigate the consequence of an accident. At 1745 the ventilation was re-adjusted and temperature returned the less than 104 F. All doors were closed and the 24 hour LCO exited. The licensee will notify this incident to the NRC Resident Inspector and the State. | ||
| ENS 42301 | 2 February 2006 05:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Emergency Filtration Train Fan Declared Inoperable | Train "A" of the Emergency Filtration Train (EFT) Unit, which services the control room ventilation system, tripped off line due to a low flow condition. The cause was determined to be a rip in the rubber boot at the suction of the fan, thus causing an automatic trip of the EFT system from a low flow condition through the filter where flow is sensed. Both the "A" and "B" trains were declared inoperable due to the amount of leakage the "B" EFT was having through the ripped boot in the "A" EFT, and the condition found on "B" EFT rubber boot. Upon further evaluation of the "B" EFT boot condition, the "B" train was declared operable at 03:02 CST on 02/02/06. The "A" EFT will remain in a 7 day LCO until the rubber boot is replaced. The 8 hour notification was issued due to both EFT Units being declared inoperable. The licensee notified the NRC Resident Inspector. | ||
| ENS 41897 | 5 August 2005 22:54:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Loss of Both Control Room Emergency Ventilation Systems During Testing | During maintenance testing of the 'B' Control Room Emergency Ventilation system (CRV) the 'A' Air conditioning unit was running. The 'A' A/C unit (V-EAC-14A) tripped on a low service water flow condition while performing step 0255-11-III-4 which manipulated service water valves for the test. The 'B' loop of the CRV system was in a 30 day LCO due to the maintenance testing. When the 'A' A/C unit tripped and the 'B' unit inoperable due to testing, both CRV trains were inoperable and that placed the unit in a 24 hour Tech Spec LCO, 3.17.A.3.a. The compressor unit V-EAC-14A was reset following restart of the 13 Emergency Service Water pump 8 minutes after the trip and exited the 24 hour LCO. The licensee notified the NRC Resident Inspector.
The notification was initiated due to both trains of the Control Room Ventilation system being inoperable and was reported under 50.72 (b)(v)(D) as an event or condition that could have prevented fulfillment of a safety function. Monticello is retracting the event notification based on further investigation of the event. Successful completion of subsequent testing indicated that the 'B' train was still capable of performing its required safety function when the 'A' train tripped. Therefore Monticello has determined there was no loss of safety function as reported in Event Notification #41897. Additional investigation is ongoing and any identified issues will be entered into the station's corrective action program. The licensee notified the NRC Resident Inspector. Notified R3DO(S. Burgess). | Service water Control Room Emergency Ventilation | |
| ENS 41567 | 5 April 2005 22:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Vulnerability with Alternate Shutdown System (Asds) Isolation Design | The following information was obtained from the licensee via facsimile (licensee text in quotes): On April 5, 2005 at 1600 (hrs. CDT), Monticello Nuclear Generating Plant during a review of the Alternate Shutdown System (ASDS) as part of the corrective actions for LER 2005-01, submitted on April 4, 2005 (Event Notification #41436) discovered a second breaker affected by a similar cause as identified in the LER. The Bus 16 source (Breaker 152-609) to Load Center #104 has a similar potential vulnerability with the ASDS isolation design that could result in Load Center #104 being locked out in the event of a Control Room or Cable Spreading Room fire. The Monticello Appendix R Safe Shutdown Analysis for Control Room/Cable Spreading (Room) fire assumes a loss of control of Division I and II equipment from the Control room, however safe shutdown is achieved remotely from the ASDS panel. ASDS design is such that a Control Room/Cable Spreading Room fire would not impede the ability to safely shutdown and maintain the plant in a shutdown condition. Contrary to the ASDS design, it was discovered that an unisolated metering circuit could result in Load Center #104 being locked out in the event of a Control Room/Cable Spreading Room fire. The bus lockout is not isolated by the ASDS transfer switches, therefore, this condition could result in failure of Load Center #104 to re-energize during the implementation of the Shutdown Outside Control Room procedure. ASDS is not required to be operable at this time. As a result of this determination, MNGP will issue a revision to LER 2005-01 to the NRC to reflect the new information. This event is being reported as a potential loss of safety function (10CFR50.72(b)(3)(v)(A,B, and D) and as a degraded or unanalyzed condition (10CFR50.72(b)(3)(ii)(B). The NRC (Resident Inspector) has been notified. | 05000263/LER-2005-001 | |
| ENS 41267 | 15 December 2004 13:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection (Hpci) System Pump Bearing Oil Plug Found Loose | The oil plug on the HPCI booster pump bearing was discovered to be loose at approximately 2100 (CST) on 12/14/2004. Upon discovery, the plug was re-tightened. The plug may or may not have fallen out had the HPCI system initiated. HPCI is operable, but subsequent evaluation has determined this to be reportable because, at the time of discovery, HPCI operation could not be assured. Event investigation is on-going. The licensee will notify the NRC Resident Inspector. | High Pressure Coolant Injection | |
| ENS 40886 | 21 July 2004 10:12:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Control Room Ventialtion (Crv) Systems Inoperable. | Both control room ventilation systems were inoperable due to a seal failure on the in service control room ventilation unit. A 24 hour Limiting Condition of Operation (LCO) was entered at 0512 CDT. (V-EAC-14A) "A" CRV tripped and "B" CRV was isolated for planned maintenance. "B" CRV (V-EAC-14B) was unisolated and restored to service at 0545 CDT (33 minutes later) and the 24 hour LCO exited at 0600 CDT. The plant remains in a 30 day LCO for one train of the CRV being inoperable. Control room temperatures increased slightly during CRV inoperability and are within normal operating band at this time (Temp increased 5 degrees F). State and Local were notified of this by the licensee. The NRC Resident Inspector was notified of this event by the licensee. | ||
| ENS 40481 | 26 January 2004 20:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 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). | Emergency Diesel Generator | |
| ENS 39970 | 30 June 2003 20:45:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | High Energy Line Break (Helb) Door Not Latched | On June 30, 2003 at approximately 1545 (CDT) it was identified that a High Energy Line Break (HELB) door separating Divisional Motor Control Centers was not latched as required. It was determined that this condition existed for a maximum of 15 minutes. This condition is being reported as an event or condition that could have prevented the fulfillment of a safety function in accordance with 10 CFR 50.72(b)(3)(v). The licensee notified the NRC Resident Inspector and the State Emergency Management Agency.
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 than is allowed by station procedural guidance. R3DO (C. Miller) notified. The station continues to review the events in the station's corrective action program. The NRC Resident Inspector was notified of this retraction by the licensee. |