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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57199 | 28 June 2024 06:10:00 | 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 | Both Divisions of Lpci Inoperable | The following information was provided by the licensee via email: This condition is being reported in accordance with 10 CFR50.72(b)(3)(v) as a condition that could have prevented fulfillment of a safety function. On 6/27/2024 at 2158 CDT, (technical specification) TS 3.5.1 condition 'D' (both divisions of (low pressure coolant injection) LPCI inoperable) was entered for surveillance testing. On 6/28/2024 at 0110 CDT, MO-2012 (residual heat removal) RHR Division 1 LPCI injection outboard valve was attempted to be cycled. It was discovered to be inoperable resulting in an inability to exit TS 3.5.1 'D'. Initial review of this condition for immediate reportability under 50.72(b)(3)(v) event or a condition that could have prevented fulfillment of a safety function, concluded the condition was not reportable based on the operability of other emergency core cooling systems (ECCS). Specifically, core spray and high pressure coolant injection were both operable to perform the function of emergency core cooling. Subsequent reviews determined that the reportability decision under 50.72(b)(3)(v) as an event or condition that could have prevented fulfillment of a safety function should be based on the safety function at the LPCI system level, rather than at the ECCS system level. The decision to report the inoperability of LPCI under 50.72(b)(3)(v) was made at 1030 CDT on 6/28/2024. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Core Spray Emergency Core Cooling System | |
| 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 52154 | 5 August 2016 03:40:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Fire Barrier | At 2240 CDT on August 4, 2016, it was discovered that the floor between the cable spreading room and the plant administration building (PAB) basement is not a credited Appendix R fire barrier. Because the cable spreading room and the plant administration building are located in the same fire area, a fire in the PAB could spread to the cable spreading room requiring evacuation of the control room. The travel path used to access the Alternate Shutdown Panel following control room evacuation traverses the same fire area in the PAB. Therefore, this event is being reported under 10 CFR 50.72(b)(3)(ii) for Degraded or Unanalyzed Condition as a fire in the PAB could have the potential to impact Division 1 equipment as well as impede the Operators ability to access Division 2 safe shutdown equipment. Fire watches have been established. There is no impact to the health and safety of the public. The NRC Resident Inspector has been notified. The licensee will notify the State of Minnesota. | 05000263/LER-2016-002 | |
| 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 50806 | 10 February 2015 18:40:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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) | 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 50663 | 10 December 2014 00:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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). | ||
| 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 50456 | 14 September 2014 07:26:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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). | ||
| 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 50117 | 15 May 2014 21:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Degraded Fire Barrier | A degraded fire barrier was identified during a walkdown on May 15th, 2014 at 1620 (CDT). The barrier separates Appendix R safe shutdown divisional equipment. A fire watch has been established as a compensatory measure. The health and safety of the public was not jeopardized as a result of this condition as there have been no fires or initiating events. The discovery of this degraded fire barrier is being reported as an unanalyzed condition as defined by 50.72(b)(3)(ii)(B). The licensee has notified the NRC Senior Resident Inspector. | ||
| 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 49938 | 20 March 2014 15:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Appendix R Fire Door Failed to Close and Latch | At 1020 CDT, door 410B did not automatically close and latch as required. Door 410 B is an Appendix R fire door that is required for divisional separation of safe shutdown equipment. Due to the doors inability to close and latch as required, divisional separation could not be assured in the event of a fire. A continuous fire watch was established once the deficiency was discovered. The door was repaired and verified to be working properly. The door was non-functional for approximately one hour and fifteen minutes from the time of discovery. Health and safety of the public was maintained as the plant was in a normal condition and there has been no actual condition needing the door to close and latch. The NRC Resident Inspector has been notified. | 05000263/LER-2014-005 | |
| 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 49814 | 10 February 2014 22:50:00 | 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)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Single Failure Vulnerability Affecting Emergency Diesel Generator Operability | The licensee also reported an additional 8 hour Non-emergency report in accordance with 10CFR50.72(b)(3)(v)(D) On February 10, 2014, (at 1650 CST) Monticello station personnel identified a vulnerability where a single failure could result in the Emergency Diesel Generators (EDGs) picking up load on the essential busses in a time frame longer than what is required by Monticello Technical Specification Surveillance Requirement (TS SR) 3.8.1.12. This surveillance requires that on a simulated or actual loss of off-site power signal in conjunction with an actual or simulated ECCS initiation signal, the Emergency Diesel Generators auto-start from standby condition and energize permanently connected loads in approximately 10 seconds. The single failure vulnerability could result in the EDGs energizing the connected loads in a slightly longer time period based upon actual test data (< 11 seconds). As a result, Technical Specification SR 3.8.1.12 was declared not met and both EDGs were declared inoperable. Monticello has subsequently isolated the single failure vulnerability and declared the EDGs operable. The station remained in a safe condition during the discovery of this vulnerability. Both EDGs remain available and functional (and operable), off-site power remains available, and the plant continues to operate in a normal condition with no initiating events present. The single failure vulnerability was associated with the 1AR 13.8KV transformer logic. The 1AR transformer is one of three off-site power sources. 1R and 2R meet the TS requirements for Off-site power. The licensee notified the NRC Resident Inspector and the State of Minnesota Duty Officer. | Emergency Diesel Generator | 05000263/LER-2014-004 |
| 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 49335 | 9 September 2013 19:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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. | ||
| ENS 49316 | 1 September 2013 21:10:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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). | Low Pressure Coolant Injection | |
| ENS 49314 | 29 August 2013 16:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Failure to Address All the Effects of External Flooding Scenarios | On August 28, 2013, Monticello Nuclear Generating Plant (MNGP) was notified of the NRC's final significance determination for a finding involving the failure to maintain a procedure addressing all of the effects of an external flooding scenario on the plant. Specifically, MNGP failed to maintain flood Procedure A.6, 'Acts of Nature,' such that it could support the timely implementation of flood protection activities within the 12 day timeframe credited in the design basis as stated in the updated safety analysis report. The finding is not a current safety concern. On February 15, 2013, actions were completed to reduce the flood mitigation plan timeline to less than 12 days by developing an alternate plan for flood protection features, pre-staging equipment and materials, improving the quality of the A.6 procedure, and preplanning work orders necessary to carry out Procedure A.6 actions. The NRC Resident Inspector has been notified. | ||
| ENS 49312 | 28 August 2013 02:52:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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). | Low Pressure Coolant Injection | |
| ENS 49293 | 22 August 2013 01:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition for Emergency Diesel Generator Fuel Oil Pumps Train Separation | On August 21, 2013 at 2000 CDT, it was determined following receipt and review of an NRC position document, that the design of the Monticello Nuclear Generating Plant diesel fuel oil supply system is not consistent with current and historical licensing and design basis documents. This condition affects fuel oil supply from the diesel fuel oil storage tank to both emergency diesel generators and is being reported pursuant to 10CFR50.72(b)(3)(ii) as an unanalyzed condition that could significantly degrade plant safety. Actions are in progress to address the unanalyzed condition. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| 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 49085 | 31 May 2013 19:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Concerning Flooding Mitigation | On May 31, 2013, during an aggregate review of issues raised during a focused self assessment of external flooding mitigation, it was concluded that the A.6 Acts of Nature procedure may not adequately protect equipment required to maintain safe shutdown from the external probable maximum flood. The plant is currently in Mode 4, Cold Shutdown for refueling. MNGP (Monticello Nuclear Generating Plant) is addressing the inadequacies. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) Unanalyzed Condition. Compensatory measures are being prepared including procedure changes, additional barriers, and contingency actions. The licensee notified the NRC Resident Inspector. | ||
| ENS 48919 | 13 April 2013 03:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Isolated Differential Pressure Switch on Safety Relief Valve Tailpipe | On April 12th, 2013 at 2200 CDT, investigation into the inability to complete a Safety/Relief Valve (S/RV) discharge line excess flow check valve test determined that a relief valve discharge monitoring instrument valve had been inappropriately closed since late June, 2011. The closed valve isolated two differential pressure switches that impact the operation of 'E' Low-Low Set (LLS) valve. The LLS logic and instrumentation were affected by the loss of two 'E' S/RV tailpipe discharge pressure switches which indicate S/RV open status and start two inhibit timers which prevent plant operators or the LLS S/RV logic from immediately re-opening the valve to allow the water leg in the S/RV discharge line to recede. The LLS logic and instrumentation is designed to mitigate the effects of postulated thrust loads on the S/RV discharge lines by preventing subsequent actuations with an elevated water leg in the S/RV discharge line. It also mitigates the effects of postulated pressure loads on the torus shell or suppression pool by preventing multiple actuations in rapid succession of the S/RVs subsequent to their initial actuation. The valve found closed has been returned to its normal open position. This condition resulted in the 'E' S/RV LLS function being aligned contrary to its design configuration and as such is being reported as an unanalyzed condition as defined by 10 CFR 50.72(b)(3)(ii)(B). The licensee has notified the NRC Resident Inspector. | Safety Relief Valve | |
| ENS 48616 | 20 December 2012 22:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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). | ||
| 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 48072 | 5 July 2012 17:58:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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). | ||
| ENS 47909 | 8 May 2012 14:00:00 | 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)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | an Unanalyzed Condition Could Delay Transfer to Emergency Diesel Generators Under Certain Postulated Conditions | At 0900 (CDT) on May 8, 2012, it was determined that Monticello Nuclear Generating Plant did not meet Technical Specification Limiting Condition for Operation 3.3.8.1 because the requirement of Table 3.3.8.1-1 for the 4.16 KV Essential Bus Degraded Voltage time delay transfer to the Emergency Diesel Generators (EDGs) of 9.2 seconds could not be met under all postulated conditions. The degraded transfer scheme has the ability to transfer to an intermediate offsite source (1AR) which under a degraded voltage condition, coincident with an accident, would delay the transfer to the EDGs an additional 5 seconds. Both EDGs were subsequently declared inoperable. As an interim action 1AR transformer has been removed from service. This eliminates the unanalyzed condition, restores Technical Specification compliance, and restores both EDGs to an operable condition. Additionally this event is being reported under criteria 10 CFR 50.72(b)(3)(v)(D) - Accident Mitigation. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| 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 47237 | 2 September 2011 21:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Intake Structure Fire Suppression System Failed Surveillance Test | During intake structure fire suppression sprinkler system surveillance testing, Operations identified that a portion of the sprinkler system was not able to pass flow. A 14-day fire protection system impairment was entered and a continuous compensatory fire watch with backup suppression was stationed prior to removing the system from service for testing. Upon failure of the surveillance test, the impairment remained in effect and the continuous fire watch remained stationed pending investigation/repair by Maintenance. On 9/2/2011 at approximately 1600 (CDT), Mechanical Maintenance personnel informed Operations that the sprinkler suppression piping was found to be fouled and not capable of passing flow. The intake sprinkler system is relied upon in part to satisfy an exemption for the station to 10CFR50 Appendix R, Section III.G.2.B concerning separation of components in the intake structure. Based upon the intake sprinkler system being non-functional, this condition is being reported under 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition affecting plant safety systems. The licensee informed the NRC Resident Inspector. | ||
| 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 46417 | 12 November 2010 18:10:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed 10Cfr50 Appendix R Scenario | At approximately 1210 on November 12, 2010, a fire protection assessment identified a potentially unanalyzed condition in the plant's 10CFR50 Appendix R analysis. In the unlikely event of a fire in the main control room or cable spreading room, coincident with a fire induced loss of off site power, in which the control room must be evacuated, Operations personnel would proceed to the Alternate Shutdown System (ASDS) panel to perform required safe shutdown activities. During certain scenarios, High Pressure Coolant Injection (HPCI) may start on a reactor low-low water level signal or on high drywell pressure signal. Since HPCI is not controlled manually from the ASDS, a postulated fire induced short could prevent the HPCI system high reactor water level trip and possibly result in reactor vessel overfill. Operation of the Safety Relief Valves in this condition has not been analyzed. For this scenario, the following unlikely sequence of events is required: 1. A fire would have to occur in the main control room or cable spreading room 2. The fire would have to be significant enough to require main control room evacuation 3. Offsite power would have to be lost 4. HPCI would have to initiate 5. A fire induced short preventing HPCI from tripping automatically would have to occur Applicable safety systems remain operable, and Operations personnel are trained in procedures to handle complex fire scenarios, including fires in the main control room and cable spreading room. Additionally, the cable spreading room is protected by an automatic Halon fire suppression system. As a precaution, a fire watch has been established as a compensatory measure. This event is being reported under 10CFR50.72(b)(3)(ii)(B). The licensee notified the NRC Resident Inspector. | High Pressure Coolant Injection Safety Relief Valve | |
| 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 46327 | 11 October 2010 18:05:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 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) | Reactor Water Cleanup | |
| 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 |