Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| 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. | |
| ENS 40008 | 22 July 2003 14:48:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Helb Door Not Latched as Required Due to Personnel Error | A High Energy Line Break (HELB) door was not latched as required. This condition is being reported as an event that could have prevented the fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(A). The door is currently closed. The HELB door which separates two critical Motor Control Center (MCC) areas was unlatched for less than two (2) minutes. The licensee will inform the state representative and has informed the NRC resident inspector.
Because plant safety was not significantly degraded, this event is not reportable under the unanalyzed condition criteria based on: (1) the door in either event was in an uncontrolled condition for less than one minute, (2) the door was not materially affected, only operated improperly, (3) the PRA significance of the event was low, and (4) the HELB Barrier door was not open for a period longer than is allowed by station procedural guidance. R3DO (C. Miller) notified. The station continues to review the event in the station's corrective action program. The NRC Resident Inspector was notified of this retraction by the licensee. | |
| ENS 40146 | 10 September 2003 22:46:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to a Potential Break in Fire Protection Water System | Potential break of Fire Water Main in Admin Bldg could cause both divisions of Safe Shutdown equipment to be inoperable by flooding the battery rooms for #11 & 12 125VDC Batteries. Vulnerable section of piping has been isolated to eliminate the flooding concern. Fire Protection compensatory actions have been established in accordance with the site Fire Protection Program. The licensee informed the NRC resident inspector.
Monticello is retracting the event reported based on evaluations which indicate that the fire main is not considered a potential flooding source. Additional evaluations are ongoing, and issues will be entered into the station's corrective action program. The licensee has notified the NRC Resident Inspector. Notified R3DO (Hills) | |
| 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). | |
| ENS 40518 | 13 February 2004 13:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Loss of Offsite Communications | At 0730, Monticello Nuclear Generating Plant became aware of the loss of incoming phone call capability. Investigation determined loss of ENS, HPN (FTS) lines were inoperable as well as the State of MN dedicated line. Commercial incoming calls (inoperable) out calls were operable. By 0930 CST all phone system were restored and determined operable. NRC Operations Center was notified of inoperable ENS phone via backup commercial line. NRC Resident Inspector was notified of this by the licensee. | |
| ENS 40585 | 11 March 2004 20:30:00 | Other Unspec Reqmnt | Voluntary Report Involving Potential Tsc Unavailability During a Dba Loca | Monticello Nuclear Generating plant is making a voluntary report with regard to the Technical Support Center (TSC) not meeting design criteria Subsection 8.2-1.f of Supplement 1 to NUREG-0737. This specifies that the TSC will be provided with radiological protection necessary to assure that the radiation exposure to any person working in the TSC would not exceed 5 REM whole body (or its equivalent part of the body) for the duration of the accident. During review of the calculations associated with an on-going Alternative Source Term project, plant staff identified the potential for a radiation shine path to exist from the reactor building to the TSC during a DBA (Design Basis Accident) - Loss of Coolant Accident (LOCA), that could result in radiation levels reaching a point dictating evacuation of the TSC under existing emergency plan procedures. As required by NUREG-0696 and confirmed by the plant staff, existing procedural guidance directs personnel to evacuate to the back-up TSC (located in the EOF) if the TSC cannot be occupied continuously. The NRC resident has been informed of this discovery.' The licensee is continuing their assessment and will determine the appropriate corrective actions. | |
| 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 40924 | 4 August 2004 13:38:00 | 10 CFR 26.73, Applicability | Fitness for Duty | A licensed employee was determined to be under the influence of alcohol during a for cause test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | |
| ENS 41005 | 1 September 2004 17:45:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Appendix R Criteria Not Satisfied for Safe Shutdown Equipment | Unanalyzed condition due to missing fire barrier and inadequate separation involving power supply cables for RHR and Core Spray pumps. During a review of the site Appendix R program, NMC Engineering personnel discovered that the credited Monticello Division I RHR Pump and Core Spray Pump power cables pass through a Division II fire area. A fire in this room could potentially damage both divisions of post-fire safe shutdown equipment. This condition is reportable under 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety due to the lack of required separation of post-fire safe shutdown trains. As a compensatory action an hourly fire watch has been established in accordance with the Monticello Fire Protection Program. The licensee informed the NRC Resident Inspector. The licensee is continuing their evaluation to determine the extent of the condition. | |
| 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. | |
| ENS 41374 | 4 February 2005 19:37:00 | 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 | 4160 Volt Relaying and Metering Single Failure Vulnerability | The following information was obtained from the licensee via facsimile: On February 4, 2005, at 1337 hours, Monticello Nuclear Generating Plant discovered a potential vulnerability with 4160 VAC current sensing and protective relaying circuitry that could result in bus lockouts of both safeguards buses (#15 and #16) if a specific equipment fault were to occur. The 1AR Auxiliary Reserve Transformer source to each of the safeguards buses have current transformers used for over-current protective relaying that have common connections to facilitate a single watt-hour meter. The lack of neutral over-current trip relaying limits the event vulnerability to a case (most likely fire) of an outside voltage source contacting one or more CT phase legs that forces current through the over-current trip relays. If this forced current is of sufficient magnitude through both division over-current relays, both safeguards buses will receive lockout signals. This would make both safeguards buses unavailable. Since the 1AR transformer is not required at this time, it has been isolated from the safeguards buses, and their associated over-current relaying isolated to preclude occurrence of this event. This event is being reported as a potential loss of safety function (10CFR50.72(b)(3)(v)(A, B, C, and D)) and as a degraded or unanalyzed condition (10CFR50.72(b)(3)(ii)(B)), The NRC Resident Inspector has been notified. See similar events #41362 (Crystal River), #41366 (LaSalle) #41369 (Quad Cities) and #41370 (Dresden). | 05000263/LER-2005-001 |
| ENS 41436 | 23 February 2005 18:00:00 | 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)(iv)(A), System Actuation | Potential Vulnerability with Alternate Shutdown System (Asds) Isolation Design. | The licensee provided via facsimile the following report: During an extent of condition review of the corrective actions associated (with) Event Notification #41374, the Monticello Nuclear Generating Plant (MNGP) engineering staff made the following discovery. On February 22, 2005 at 12:00 hours, MNGP discovered a potential vulnerability with Alternate Shutdown System (ASDS) isolation design which could result in Bus 16 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 un-isolated metering circuit from the 1AR transformer could result in Bus 16 being locked out in the event of a Control Room or Cable Spreading Room fire. The bus lockout relay from the 1AR transformer is not isolated by the ASDS transfer switches, therefore, this condition could result in failure of Bus 16 to re-energize during the implementation of the Shutdown Outside Control Room procedure. Since the Bus 16 feeder breaker from the 1AR transformer is not required at this time, it has been isolated from the safeguards bus to preclude occurrences of this event. The event is being reported as a potential loss of safety function (10CFR50.72(b)(v)(A,B and D) and as a degraded or unanalyzed condition (10CFR50.72(b)(3)(ii)(B)). The licensee informed NRC Resident Inspector. | 05000263/LER-2005-001 |
| ENS 41441 | 25 February 2005 03:11:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Esf Actuation Following Trip of Reactor Protection Motor-Generator Set | A trip of the "A" RPS M-G set resulted in an "A" Group 2 isolation and startup of the standby gas treatment system. There was a preliminary report of a possible fire/smoke smell in the vicinity of the M-G set. However, when an operator reported to the location there was no observed fire. The fire brigade was also dispatched but found no fire or smoke in the M-G set area. The licensee is preparing to place the RPS on its alternate power supply. This will allow the Group 2 isolations and standby gas treatment system actions to be reset. The cause of the M-G set trip is still under investigation. The licensee will be notifying the NRC Resident Inspector as well as state and local authorities. | 05000263/LER-2005-002 |
| ENS 41468 | 8 March 2005 10:54:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Rhr Pump Tripped Due to Loss of Valve Position Indication | The following information was provided by the licensee (licensee text in quotes) During isolation of the 'A' Safety Relief Valve, the 12 Residual Heat Removal (RHR) Pump tripped while in service for shutdown cooling. The isolation and shutdown cooling were subsequently restored such that shutdown cooling was lost for approximately 13 minutes. Several Control Room alarms were received at approximately 0454 (CST) while the isolation was being hung. The Control Room Supervisor investigated the alarms and the Reactor Operator identified that the 12 RHR Pump had tripped. The isolation was lifted and shutdown cooling was restored at approximately 0507 (CST). Operators observed that reactor water temperature and level remained stable at 99 degrees F and 651 inches (above the bottom of the vessel), respectively, during the event. The isolation being hung apparently caused a loss of position indication for the RHR pump inlet valve. With no position indication, the pump logic sensed a loss of suction flow path, which caused the RHR pump to trip. The event is under investigation. The NRC resident has been notified. | |
| 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 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). | |
| ENS 42013 | 25 September 2005 19:10:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Degradation in Fire Barrier Discovered During an Inspection | During an inspection of structural steel in the Emergency Filtration Train (EFT) building, it was determined that a portion of the steel did not have adequate fire retardant material protecting the steel. This condition would have compromised the 3 hour-fire barrier between the 2 divisions of the Emergency Service Water (ESW) System, impacting # 11 Emergency Diesel Generator ESW Pumps, # 13 and # 14 ESW Pumps. This condition could have resulted in the inability to establish and maintain cold shutdown conditions in the event of a fire in this area. A fire impairment was entered and an hourly fire watch was established to address this condition. All ESW divisions are presently operable. The NRC Resident Inspector will be notified of this event by the licensee.
Monticello is retracting the event notification based on further investigation of the issue. The station has completed calculations that confirm the affected steel beams are not required to maintain fire barrier integrity. Further, the fire severity in the zone will not cause the steel beams to fail. Therefore, the ability of the station to establish and maintain cold shutdown conditions in the event of a fire in this area was not impacted. Based on this information, Monticello has determined there was no unanalyzed condition as reported in Event Notification # 42013. The degraded fire retardant material has been replaced. The licensee notified the NRC Resident Inspector. | |
| 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 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 42652 | 19 June 2006 04:19:00 | 10 CFR 50.73(a)(1), Submit an LER | Radiation Monitor Spiked Three Times | At 2319 on 06/18/2006, the 'A' Plenum Radiation Monitor spiked high which resulted in the closure of the Drywell CAM (Continuous Air Monitor) and the Oxygen Analyzer Primary Containment Isolation Valves. The spike's (3) of approx. 30 Mr/HR occurred during this event. The valves isolated once, the Reactor Building Ventilation trip was reset once and re-isolation occurred several minutes later from another spike on the Rad Monitor. Trip setpoint is 26 Mr/hr. The Plenum high Rad signal also resulted in Reactor Building isolation (twice), start of 'A' Standby Gas Treatment, and transfer of the Control Room Ventilation to the high Rad Mode. The 'B' Plenum Rad Monitor remained constant at 1.3 Mr/hr. The Reactor Building Ventilation and Control Room Ventilation have been reset and Standby Gas Treatment has been secured. The 'A' Plenum Radiation Monitor has been declared inoperable. Instrument & Control Technicians are currently troubleshooting the problem associated with the 'A' Plenum Radiation Monitor. The NRC Resident Inspector's have been left messages by the licensee.
This report is being reclassified by the licensee from a 50.72(b)(3)(iv)(A) (valid system actuation) to 50.73(a)(2)(iv)(A) (invalid system actuation). Based on further investigation, Monticello has determined that the actuation signal was due to a failed microswitch and therefore was not a valid signal. Because the actuation signal was not valid, Monticello is retracting the initial event report and instead reporting this event as an unplanned system actuation under 50.73(a)(2)(iv)(A). This report will be made in lieu of reporting the event as an LER. In accordance with 50.73(a)(2)(iv)(A): This report is not considered an LER and the report is being made under 50.73(a)(2)(iv)(A). The original event report (ENS #42652) detailed the systems affected, whether the actuation was complete or partial and whether each affected system started and functioned successfully. The cause of the invalid signal was the failure of the trip check pushbutton (micro)switch due to age related degradation. The switch has been replaced and the 'A' Plenum Radiation Monitor was returned to service. The licensee notified the NRC Resident Inspector. Notified the R3DO (O'Brien). | |
| ENS 42740 | 30 July 2006 22:30:00 | 10 CFR 50.73(a)(1), Submit an LER | Secondary Containment Isolation Due to 'B' Spent Fuel Rad Monitor Failure | At 1730 on 7/30/06 the 'B' Spent Fuel Pool Radiation Monitor spiked high, which resulted in the closure of the DW CAM (Continuous Air Monitor) and the Oxygen analyzer primary containment isolation valves. The spikes (2) of approximately 50 Mr/hour occurred during this event. Only 1 automatic isolation occurred. The Spent Fuel Pool Radiation Monitor High Rad signal also resulted in a Reactor Bldg Isolation (Secondary Containment), start of 'A' Standby Gas Treatment, and transfer of the Control Room Ventilation to the High Rad Mode. The 'A' Spent Fuel Pool Monitor remained constant at its background radiation level. No activities were in progress on the Refuel Floor at the time of the trips. Rad Protection Tech surveyed the area with no abnormal readings noted. The 'B' Spent Fuel Pool Radiation Monitor was declared inoperable. All automatic isolation valves have been reset, Rx Bldg Ventilation, Control Room Ventilation have been reset and SBGT (Standby Gas Treatment) has been secured. Normal trip setpoint is 50 Mr/ hour. The licensee has taken the actions of LCO 3.2.E The licensee notified the NRC Resident Inspector.
This report is being reclassified from a 50.72(b)(3)(iv)(A) (valid system actuation) to 50.73(a)(2)(iv)(A) (invalid system actuation). Based on further investigation, Monticello has determined that the actuation signal was due to a failed micro switch and therefore was not a valid signal. Because the actuation signal was not valid, Monticello is reclassifying the initial event report as an unplanned system actuation under 50.73(a)(2)(iv)(A). This report will be made in lieu of reporting the event as an LER. In accordance with 50.73(a)(2)(iv)(A): '"This report is not considered an LER and the report is being made under 50.73(a)(2)(iv)(A). The original event report (EN #42740) (see above) detailed the systems affected, whether the actuation was complete or partial, and whether each affected system started and functioned successfully. The cause of the invalid signal was the failure of the trip check pushbutton micro switch due to age related degradation. The switch has been replaced, and the 'B' Spent Fuel Pool Radiation Monitor was returned to service. The licensee has notified the NRC Resident Inspector. Notified R3DO T. Kozak. | |
| ENS 42788 | 19 August 2006 23:49:00 | 10 CFR 50.73(a)(1), Submit an LER | Radiation Monitor Electrical Spike Causes Secondary Containment Isolation | The 'B' fuel pool radiation monitor spiked high causing an actuation of the secondary containment relays. 'A' SBGTS (Standby Gas Treatment System) train started, reactor building, turbine building, and rad waste building ventilation isolated. The Drywell Continuous Air Monitor (DW CAM) and O2 analyzer containment valves isolated. The control room ventilation system transferred to the high radiation mode. At the time of the occurrence the 'A' Fuel Pool radiation monitor was reading normal (0.5 mr/hr). The 'B' Fuel Pool radiation monitor was reading 80 mr/hr. The 'B Fuel Pool radiation monitor, soon after lowered to 30-40 mr/hr and then back to normal (10 mr/hr). The trip setpoint is 50 mr/hr. The trip was reset. The secondary containment isolation and all ventilation trips were restored to normal. The DW CAM and control room ventilation were restored to normal. A local survey was performed on the refuel floor, readings obtained were < 5 mr/hr at the detector and < 2 mr/hr in surrounding areas. The 'B' fuel pool radiation monitor high level trips were placed in the trip/bypass position. The 'B' fuel pool radiation monitor was placed in downscale trip condition to comply with Technical Specification Table 3.2.4. A team is being established to investigate the cause of the trip. The licensee notified the NRC Resident Inspector.
This report is being reclassified from a 50.72 (b)(3)(iv)(A) (valid system actuation) to 50.73(a)(2)(iv)(A) (invalid system actuation). Based on further investigation, Monticello has determined that the actuation signal was an invalid spurious signal since other radiation monitors in the area did not indicate any change in value. The cause of the spurious signal was determined to be inducted noise through an unshielded cable which resulted in an increased output current to the instrument. Because the actuation signal was not valid, Monticello is reclassifying the initial event report as an unplanned system actuation under 50.73(a)(2)(iv)(A). This report will be made in lieu of reporting the event as an LER. In accordance with 50.73(a)(2)(iv)(A): This report is not considered an LER and the report is being made under 50.73(a)(2)(iv)(A). The original event report (ENS #42788) detailed the system affected, whether the actuation was complete or partial, and whether each affected system started and functioned successfully. The licensee has notified the NRC Resident Inspector.' NRC R3DO (Julio Lara) notified. | |
| ENS 43088 | 10 January 2007 21:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram with Stuck Open Main Turbine Control Valves | At 1528 hours Central Time, a RPS trip occurred, all rods fully inserted. Turbine valve testing was in progress. A group 1 isolation occurred with all MSIVs closing. A partial group 2 occurred due to low reactor water level. All safety systems (valves) operated correctly. Currently reactor pressure/cooldown in progress using HPCI in pressure control, normal condensate and feedwater in service for reactor water level control. Investigation of cause is in progress. Group isolations resulted from low pressure and low reactor water level. During the transient and subsequent cooldown, Operators manually opened the main steam relief valves to maintain pressure control in accordance with the plant EOPs. HPCI was manually started to maintain reactor level. The group 2 isolation has been reset. The group 1 isolation (MSIVs) will not be reset until after the main turbine control valves are shut. The plant is being cooled down to initiate RHR cooling and is in the normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector. | |
| ENS 43095 | 11 January 2007 23:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to State Agencies | Notification to NRC following notification to the following government agencies: State of Minnesota Pollution Control Agency State of Minnesota Dept of Natural Resources Enforcement State of Minnesota Dept of Natural Resources Area Fisheries Office State of Minnesota Dept of Natural Resources Local Conservation Office Notifications made to above government agencies in accordance with Monticello Nuclear Plant water appropriations permit for fish kill in Mississippi river following Rx scram on 1-10-07. The licensee will notify the NRC Resident Inspector. | |
| ENS 43114 | 23 January 2007 19:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Press Release Issued | The purpose of this notification is to inform the NRC that Nuclear Management Company (NMC) will be issuing a press release approximately two hours (CST) after this notification to the NRC on January 23, 2007, concerning an event previously reported to the NRC on January 10, 2007, via EN# 43088. The event in question involved an automatic reactor scram at 3:26 PM on January 10, 2007. As reported in that notification, all safety systems operated correctly. The scram occurred following the unexpected opening of the main turbine control valves. There was no release of radioactivity during the event. The purpose of the press release is to provide information to the media and the public regarding the results of NMC's investigation as to the cause of the January 10 event and the status of remedial actions. The licensee notified the NRC Resident Inspector. The licensee will notify State and Local authorities. | |
| ENS 43174 | 20 February 2007 14:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite State Notification Required by Npdes Permit | Notification to NRC following notification to the following government agencies on 02/20/2007 at 0830 (CST): State of Minnesota Pollution Control Agency - Water Control Quality, State of Minnesota Department of Emergency Management - State Duty Officer. Notifications made to above government agencies in accordance with Monticello NPDES permit for a collapse of cooling tower panels which resulted in a diversion of circulating flow overland to the discharge canal, resulting in washing of soil and gravel into the canal. The licensee informed the NRC Resident Inspector. | |
| ENS 43246 | 18 March 2007 04:46:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Trip Signal and Containment Isolations Received During Maintenance | With the plant in refueling outage with all rods in and in Mode 5, a RPS trip and a containment isolation was received at 2347, 03/17/07. This event occurred when Div #2 Bus 16 Bus Pot Drawer was opened during isolation of 1AR Transformer Metering under C/O-17605. The opening of this Pot Drawer caused Div #2 4KV Bus 16, Div #2 480V LC-104 and Div #2 480V MCC-141, 142, 143, and 144 to open. 'B' RPS tripped causing full RPS trip due to SRM shorting removed. The loss of power caused RBV (Reactor Building Ventilation) and Spent Fuel Pool Radiation Monitors to trip and cause containment isolation to occur. Investigation into C/O-17605, restoration of power, resetting of RPS and containment isolation in progress. The containment valves that received an isolation signal were: Primary Containment Atmosphere Control, Post Accident H2/O2 Control system, Post Accident Sampling system, O2 analyzing and SCTMT isolation (H&V). The licensee notified the NRC Resident Inspector. | |
| ENS 43488 | 12 July 2007 21:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unallowable Manual Actions Credited for Fire Safe Shutdown | During performance of NFPA-805, Transition Project Task SUP-1, 'Manual Action Compliance,' it was determined manual operator actions are being credited in IX/12-A Lower 4KV Room and XII/14-A Upper 4KV Room to achieve and maintain hot safe shutdown. The manual actions are to provide temporary ventilation capability to the 4KV rooms to assume continued operability of vital switchgear. The switchgear provides power to equipment needed to achieve and maintain hot shutdown. These manual actions were specified in an Appendix R Section III.G.1/G.2 fire area; however, they do not meet the criteria for allowable manual actions specified in RIS 2006-10, 'Regulatory Expectations with Appendix R paragraph III.G.2 Operator Manual Actions.' No system actuations occurred as part of this event. The discovery of these manual actions is being reported as an unanalyzed condition as defined by 10 CFR 50.72(b)(3)(ii)(B) and has been entered into the site's corrective action program The alternate compensatory measure for these areas is to perform the specified manual actions. An extent of condition review will be initiated that will encompass the remainder of the safe shutdown areas. The results of the extent of condition will be documented in the site's corrective action program with compensatory measures being established as appropriate. The 60 day licensee event report, submitted to the Commission in accordance with 10 CFR 50.73(a)(2)(ii), will provide the results of the manual action compliance review and follow-up corrective actions. This is based on preliminary data and further investigation is ongoing. The licensee will notify the NRC Resident Inspector. | |
| ENS 43492 | 17 July 2007 12:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Inoperable Due to Planned Relocation | On 07/17/07 at 0700 hours CDT the Monticello Nuclear Generating Plant's (MNGP) Technical Support Center (TSC) will begin relocation to a new facility. The relocation activities include implementation of compensatory measures to maintain the TSC functions during the transition. The compensatory measures include having the Emergency Director report to the Control Room and co-locating the remaining TSC staff at the EOF should an event declared requiring Emergency Response Organization (ERO) activation. The ERO has previously successfully demonstrated the ability to implement these compensatory measures. The relocation and testing of equipment in the new TSC are scheduled to be complete(d) on or before 07/23/07. The site Emergency Response Organization has been notified of the relocation and instructed on the planned compensatory measures to be implemented during the move. MNGP will notify the NRC upon completion of the relocation and declaration of TSC operability in the new location. This event is considered reportable per 10 CFR 50.72 (b)(3)(xiii). The licensee has notified the NRC Resident Inspector.
On 07/19/07 at 1400 hours CDT the Monticello Nuclear Generating Plant's Technical Support Center (TSC) was declared operable after its relocation to a new facility. Relocation and testing activities have been completed and the Emergency Response Organization has been notified of the cessation of compensatory measures. NRC R3DO (Eric Duncan) notified. The licensee has notified the NRC Resident Inspector. | |
| ENS 43525 | 26 July 2007 14:02:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition - Flood Doors Closed Causing a Potential Loss of Both Vital Switchgear | At 09:02 on 07/28/07, an outplant 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. At 09:55 on 07/26/07, 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. The licensee notified the NRC Resident Inspector. | |
| ENS 43581 | 20 August 2007 15:53:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Electrical Breaker B4305 Unable to Function | During panel walk downs in the Main Control Room, an Operator discovered no light indication for V-AC-4, B RHR ROOM AIR COOLING UNIT. Operator was dispatched to Breaker B4305 and 'B' RHR room to determine cause of loss of indicating light. The Operator determined that breaker B4305 (V-AC-4, 'B' RHR ROOM AIR COOLING UNIT 480V SUPPLY) was unable to function. Division II RHR pumps, Division II CS pump, and associated low pressure injection systems were declared inoperable. Tech Spec Actions 3.5.1.A, 3.5.1.B, and 3.5.1.M were entered resulting in entry into LCO 3.0.3 at 1053 on August 20, 2007. At 1151 on August 20, 2007 power reduction was commenced using normal shutdown procedures. Site Electricians determined that a line fuse had blown. The fan motor was meggared and all three line fuses were replaced. At 1230 on August 20, 2007, V-AC-4 was tested and returned to service. Division II RHR pumps, Division II CS pump and associated low pressure injection systems were declared operable. LCO 3.0.3 and T.S. Actions 3.5.1.A, 3.5.1.B, and 3.5.1.M were exited. Power reduction was terminated at 98%. At the completion of the call the power level had returned to 100%. The licensee notified the NRC Resident Inspector. | |
| ENS 43788 | 16 November 2007 17:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Degredation of Emergency Preparedness Response Capabilities | At approximately 1100 CST on Friday November 16, 2007, Monticello Nuclear Generating Plant determined that there had been a degradation of the Emergency Preparedness response capabilities when the Emergency Response Data System, Offsite Dose Projection (MIDAS), Safety Data Parameter System, selected telephones, and site computer terminals became inoperable in the Technical Support Center (TSC). The loss of these systems is associated with a partial loss of power to the TSC. As a result of these conditions, the TSC was declared non-functional. Emergency Communications remain available with the Emergency Telecommunication System, and Health Physics Network being operable. Monticello has verified operability of the ERO notification system (pagers), and the communication circuit used to notify the state and local counties. The siren system for local counties was unaffected by loss of site systems. Plant page and radio systems remain operable. Adequate communications capabilities are operable at this time to implement the emergency plan. The site's Emergency Operations Facility (EOF), which is the back-up facility for the TSC, is fully operational and measures are being implemented to direct TSC personnel to the near site EOF in the event of an actual emergency condition. The station has informed Wright and Sherburne counties and the Minnesota State Duty Officer of this condition. Troubleshooting efforts are in progress. A return to service time for all systems is to be determined. Monticello has determined that this event is reportable to the NRC as an 8-hour non-emergency report in accordance with 10 CFR 50.72 (b)(3)(xiii). The licensee notified the NRC Resident Inspector. * * * UPDATE FROM K. HAUGEN TO P. SNYDER AT 1253 ON 11/17/07 * * * The TSC has been restored to an operable configuration. The licensee notified the NRC Resident Inspector. Notified R3DO (Passehl). | |
| ENS 43836 | 11 December 2007 14:55:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Loss of Rhr Room Cooler | At 08:55 on 12/11/07 the division 1 RHR room cooler, V-AC-5 would not start. At approximately 10:00, troubleshooting determined that the cause of the failure was a blown line fuse on the B phase of the 480 VAC supply breaker. In accordance with plant procedures, a loss of this room cooler requires that the associated division 1 core spray pump and both division 1 LPCI pumps be declared inoperable. The result is entry into Technical Specification 3.5.1 Condition M due to having two or more ECCS injection/spray subsystems inoperable. This requires entry into LCO 3.0.3. At 10:53 on 12/11/07 the blown line fuse was replaced, the unit was tested and operability was restored. LCO 3.0.3 was exited at this time. The Technical Specification bases for Technical Specification 3.5.1 Condition M states that when multiple ECCS subsystems are inoperable, as stated in Condition M, the Plant is in a condition outside of the accident analyses. As described in this bases section the plant is in an unanalyzed condition and pursuant to 10 CFR 50.72(b)(3)(ii) an 8 hour report is being made. At this time, the station believes there was no loss of safety function. Further review of the event is in progress at the station. The station has informed the NRC resident inspector of this event.
After further review, the licensee is retracting this report based on the following: The notification was initiated due a TS Bases that stated the plant was outside of its accident analyses with multiple ECCS subsystems are inoperable. This was considered an unanalyzed condition. Further evaluation by plant staff has determined that for the conditions present at the time of the event the station was bounded by the accident analysis and therefore the event was not an unanalyzed condition. A review of Section 14 of the Updated Safety Analysis Report (USAR) determined that with one low-pressure ECCS division inoperable, the plant was not outside the ECCS accident analysis as described in the SAFER GESTR ECCS licensing topical report for Monticello and reflected in Chapter 14 of the USAR. The licensee informed the NRC Resident Inspector. Notified R3DO (Tom Kozak). | |
| ENS 43847 | 17 December 2007 08:03:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Assessment Capability (Spds) | At 0203 CST, the plant experienced a loss of the Safety Parameter Display System (SPDS) and other computer systems used for emergency assessment in the Control Room, Technical Support Center (TSC) and Emergency Operating Facility. This capability was fully restored at 0600 CST. The cause of the loss was an electrical fault in an intercom box in the TSC. The licensee will notify the NRC Resident Inspector. | |
| 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). | |
| ENS 44211 | 3 April 2008 11:25:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of a Sbgt During Annunciator Test | Monticello Nuclear Generating Plant is making a telephone report in accordance with 10 CFR 50.73(a)(2)(iv)(A) Invalid Partial Actuation of the Standby Gas Treatment and Secondary Containment Isolation Systems due to an inappropriate operator action. This report is being made in lieu of a written Licensee Event Report. At 0825 on 04/03/2008, an operator was performing an annunciator test of the Control Room panels and inappropriately pushed the 'A' Standby Gas Treatment system 'Test' pushbutton instead of the 'Lamp Test' pushbutton. He then immediately pushed the 'Reset' pushbutton which reset the Standby Gas Treatment train. The inappropriate actuation of the 'A' Standby Gas Treatment System resulted in the 'A' train momentarily starting, causing ventilation fans V-EF-10 and V-MZ-6 to trip. The immediate resetting of the system by depressing the 'Reset' pushbutton prevented a full secondary containment isolation. All systems started and functioned successfully. The cause of the invalid signal was the operator actuating the system by depressing the 'Test' pushbutton instead of the 'Lamp Test' pushbutton. The NRC Resident Inspector was notified of this event report. After the invalid actuation was completed, all systems affected were reset and returned to normal lineup. | |
| ENS 44484 | 12 September 2008 03:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram and Isolation Due to Transformer Lockout | While the 1R transformer was out of service for maintenance, the 2R transformer experienced a lockout resulting in a loss of normal offsite power, a reactor scram, and a Group 1, 2 and 3 isolation. All rods fully inserted as expected. The cause of the 2R transformer lockout is unknown at this time. After initiation, the high pressure coolant injection (HPCI) system would not trip at the high reactor water level set point +48", as required. The operators then manually isolated the HPCI steam lines. Plant decay heat removal is with the reactor core isolation cooling (RCIC) system and the safety relief valves. Torus cooling is in service. The vital electrical busses are being supplied by the 1AR transformer. Efforts are underway to restore the 1R transformer to service, and subsequently the non-vital busses. The licensee notified the NRC Resident Inspector.
The HPCI system was declared inoperable and isolated due to failure to automatically trip at +48" Reactor Water Level. The HPCI steam supply valves were automatically closed to remove HPCI from service. In addition, the HPCI turbine trip failed to trip with the turbine trip push button. The cause of the trip failure is unknown at this time. The licensee is continuing to investigate. HPCI did automatically start as designed and injected to the reactor vessel as designed. However, HPCI failed to trip at High Reactor Water Level as required. Additionally, the Automatic Depressurization System (ADS) timer showed erratic indication following the event. The ADS timer was inhibited to prevent automatic action. ADS is inoperable, but manual steam relief valve operation remains available. The licensee will notify the NRC Resident Inspector. Notified R3DO (Passehl), NRR EO (Ross-Lee), and IRD (McMurtray).
A (second) Group 2 isolation signal was received when reactor water level lowered below +9 (inches) (while pumping drywell sumps). All Group 2 valves except the drywell sump isolation valves were closed due to a previously reported Group 2 signal. The drywell sump valves had been opened to allow manual pumpdown of the sumps. The sump valves closed as expected. Licensee notified the NRC Resident Inspector. Notified R3DO (Passehl). | 05000263/LER-2008-007 |
| ENS 44498 | 17 September 2008 15:17:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Due to Loss of Shutdown Cooling | Monticello lost the 1R offsite power transformer due to an industrial accident that grounded the bus and resulted in a fatality of an employee onsite. At the time of this event, the 2R offsite transformer was already out of service due to a previous event (see EN #44484). This effectively resulted in a loss of all normal offsite power. The 1AR safety related offsite power source remained available and the safety related buses are energized and diesels are available. However, shutdown cooling was lost during the event due to a Group 2 isolation and cannot be restored. The shutdown cooling suction isolation valves #2020 and #2030 have power but until the Group 2 logic can be reset the valves cannot be reopened. The licensee needs to repower the RPS bus to reset the logic. Based on the uncertainty in reestablishing shutdown cooling an Unusual Event was declared based on Emergency Director judgment (EAL HU5.1). Current reactor coolant temperature is approximately 110 degrees with a heatup rate of about 20 degrees an hour. State and local authorities and the NRC resident inspector have been notified. The licensee will likely be making a press release.
On Wednesday, September 17, 2008 Monticello Nuclear Generating Plant (MNGP) experienced a loss of power to the station transformer resulting in a valid actuation of the following systems: Reactor Protection System (with the reactor shutdown), Containment Isolation, and Emergency Diesel Generators. The cause of the loss of power was due to contact of a 115kV transmission line by a manlift. A vendor employee was electrocuted. On-site Medical Emergency Response personnel responded until the individual was transported to North Memorial Medical Center. The individual was pronounced dead at North Memorial Medical Center. Notifications of offsite agencies and a media press release are in progress. 'This notification is being made in accordance 10CFR50.72(b)(2)(xi) and 10CFR50.72(b)(3)(iv)(A).' Licensee notified the NRC Resident Inspector." Notified R3IRC (Garza)
This is a follow-up to Event notification #44498. The station has completed notifications to off-site agencies and to the media. The station continues to troubleshoot and restore plant equipment and respond to media inquiries. NUE update - As of 1113 CDT shutdown cooling was restored." Notified R3 IRC.
Monticello Nuclear Generating Plant exited their Notification of Unusual Event at 1100 CDT on 9/21/08. The exit criteria is supported on the following information: (1) This event did not result in the loss or potential loss of a fission product barrier and did not change the status of the current fuel condition. All three fission barriers are intact and were maintained throughout the event. (2) There were no radiation releases as a result of the event. (3) Restoration of plant loads including shutdown cooling onto a normal offsite power source (the 1R transformer) has been completed. (4) The 1R transformer offsite power source was recovered and is being protected. Protection of the 1R transformer will continue until availability of the other offsite power source supplied from the 2RS/2R transformers has been restored. (5) Shutdown cooling is in service. (6) The site organization challenges in response to the injury event and loss of decay heat removal are no longer present. The licensee has notified the NRC Resident Inspector, as well as state and local agencies. A media release will be issued. Notified R3DO (Stone), NRR EO (Galloway), IRD MOC (Clark), OPA (Burnell), DHS (S. Moore), and FEMA (D. Fuller). | 05000263/LER-2008-007 |
| ENS 44508 | 21 September 2008 02:35:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Loss of Shutdown Cooling Due to Containment Isolation | On Saturday, September 20, 2008, Monticello Nuclear Generating Plant (MNGP) experienced an actuation of the following systems: Reactor Protection System (with the reactor shutdown), Containment Isolation, and Emergency Diesel Generators. The apparent cause of the actuation was a pressure pulse in the reference leg of a Reactor level instrument that resulted when a CRD (Control Rod Drive) pump was started without the reference leg backfill system isolated from the CRD system. This notification is being made in accordance with 10CFR50.72(b)(3)(iv)(A). Due to the containment isolation, shutdown cooling was lost for approximately 90 minutes. Initial reactor temperature was ~95 degrees when the isolation occurred. When shutdown cooling was restored, reactor temperature had increased to ~120 degrees. The Emergency Diesel Generators started but did not load. The diesels have been restored to normal standby status. The licensee has notified the NRC Resident Inspector. | |
| ENS 44633 | 6 November 2008 17:15:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification of a Discharge to the Mississippi River | On November 6, 2008 at 11:15, Operations was notified that results of an October 20, 2008 oil and grease weekly sample from a plant sump discharge to the Mississippi River exceeded the National Pollution Discharge Elimination System (NPDES) daily maximum allowable value of 15ppm. The October monthly average oil and grease NPDES of 10ppm was also exceeded. The Minnesota Pollution Control Agency is being notified. The licensee notified the NRC Resident Inspector. | |
| ENS 44723 | 15 December 2008 22:16:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Steam Chase Temperature Greater than 165F | At 1616 on 12/15/08, a plant heating boiler trip resulted in a loss of a reactor building ventilation. The loss of reactor building ventilation resulted in maximum average main steam chase temperatures greater than or equal to 165F. High energy line break (HELB) analysis of piping in the steam chase assumes an initial average temperature prior to the break of 165F. Temperature greater than or equal to 165F in the steam chase challenges EQ qualification of the piping analysis. Abnormal procedures for loss heating boiler and ventilation system failure were entered. C.3 (Shutdown) and C.5-1300 (secondary containment control) were also entered. The plant heating boiler was restarted and ventilation restored prior to power reduction. All systems have been returned to normal. The licensee notified the NRC Resident Inspector.
The licensee is retracting this report based on the following: Monticello is retracting the event reported based on further evaluation, which found that the issue was not an unanalyzed condition that seriously degraded plant safety. The investigation of the event found the peak temperature achieved was 167.2 degrees F and the condition lasted for approximately 11 minutes. Engineering review of Safety System Components found no impact on the equipment for the temperature reached, Additionally, revised High Energy Line Break (HELB) calculations performed with an initial average Steam Chase Room temperature of 180 degrees F before a HELB determined that Safety System components could perform their safety functions. The station has identified the cause for the event and corrective actions will be tracked in the station's corrective action program. Since there was no impact on the equipment in either Environmental Qualification (EQ) or safety function, the temperature of the event was less than the revised calculation temperature, and the unanalyzed condition that existed in the initial event notification report no longer exists and did not result in a condition that seriously degraded plant safety, this event can be retracted. The licensee informed the NRC Resident Inspector. Notified R3DO (Ring). | |
| ENS 44797 | 22 January 2009 15:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Phone Communications | At approximately 0930 CST on 01/22/2009, the Monticello Nuclear Generating Plant became aware of a loss of ENS, HPN, and ERDS lines as well as a partial loss of in-coming commercial lines. Cell phone communications were not affected. The NRC Operations Center was contacted, and a back-up means of communication was established. The back-up communication methods to off-site State and Local agencies remained available. The ENS communications were conducted via commercial phone lines or a cell phone line to the NRC Operations Center. The NRC Resident Inspector, NRC Region III, NRC Operations Center, State and local government agencies were notified. At 1300, all communications were re-established. At 1455 communication capabilities in the Control Room, Shift Managers Office, TSC, and EOF were verified to be operating properly. Communications were lost due to a problem at the local telephone company office, and not due to any on-site equipment issues. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii). | |
| ENS 44899 | 20 January 2009 16:10:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Secondary Containment | Monticello Nuclear Generating Plant is making a telephone Report in Accordance with 10 CFR 50.73(a)(2)(iv)(A) Invalid Actuation of the Standby Gas Treatment and Secondary Containment Isolation Systems due to high resistance resulting from inadequate contact wipe of a relay. This report is being made in lieu of a written Licensee Event Report. SPECIFIC TRAINS AND SYSTEMS THAT WERE ACTUATED The 'A' Standby Gas Treatment system actuated and Secondary Containment isolated. DESCRIPTION OF WHETHER EACH TRAIN ACTUATION WAS COMPLETE OR PARTIAL On January 20, 2009, during performance of Step 29 of Procedure 0003, Drywell High Pressure Scram and Groups 2, 3, & Secondary Containment Isolation Test and Calibration, the Standby Gas System automatically initiated and Secondary Containment isolated due to high contact resistance on a Drywell High Pressure HFA relay. This resulted in increased steam chase temperatures, entry into Action Statement TS 3.3.6.2.A for Secondary Containment Isolation Instrumentation, entry into Action Statement TS 3.6.4.1.A for Secondary Containment, entry into Action Statement TS 3.6.4.3.A for Standby Gas Treatment, and 10CFR50.73 reportability. DESCRIPTION OF WHETHER OR NOT THE SYSTEM STARTED AND FUNCTIONED SUCCESSFULLY All systems started and functioned successfully. The cause of the initiation of SBGT and isolation of Secondary Containment during performance of Step 29 of Procedure 0003 was the presence of high resistance at relay 16A-K60A, contacts 5-6. The cause of high resistance at relay 16A-K60A, contacts 5-6, was inadequate contact wipe. The cause of the inadequate contact wipe was inadequate contact wipe adjustment after replacement of HFA coils with the new Century series coil. The NRC Resident Inspector was notified of this event report. | |
| ENS 44907 | 16 March 2009 14:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Fish Kill During Routine Shutdown | The licensee notified state and local authorities of a fish kill that occurred during a routine plant shutdown for refueling. The fish kill, in the Mississippi River, was monitored for a two day period. Notifications were made to the State of Minnesota Emergency Management Agency, Wright County Sheriff's Office, and Sherburne County Sheriff's Office. The licensee has notified the NRC Resident Inspector. | |
| ENS 44957 | 2 April 2009 10:43:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Residual Heat Removal Inoperable | At 0543 (CDT) on April 2, 2009 at the Monticello Nuclear Generating Plant (MNGP) an Operator made the following discovery during performance of his rounds. The flowrate and pressure of the #14 Residual Heat Removal Service Water (RHRSW) pump motor cooling appeared to be low. Investigation found the flow to be approximately 1 gpm. The cooling water supply flow to the pump motor cooler comes from either RHRSW or Service Water. Based on these indications, Operators declared the RHR Shutdown Cooling inoperable and entered actions for Technical Specifications 3.9.7. Actions have been completed to provide an alternate water supply to the RHRSW pump motor oil cooler. Based on the inoperability of the RHR Shutdown Cooling, this event is reportable under 10 CFR 50.72(b)(3)(v), 'An Event or Condition that could have Prevented the Fulfillment of a Safety Function-Capability to Remove Residual Heat.' The station is performing further investigation into the event and will develop corrective actions based on the results of the investigation. The Licensee notified the NRC Resident Inspector.
Monticello is retracting the event reported based on further evaluation. An investigation of the event found test data that demonstrates the RHRSW pump would not have lost its ability to provide cooling water to the shutdown cooling system and therefore no loss of safety function occurred. The test data provides documentation that the thrust bearing oil bath temperature of the RHRSW pump motor would not have exceeded the 200 deg F limit imposed by the motor supplier (GE) at the flow rate found by the operator. The test data indicated with the cooling water at a flow rate of less than 0.9 GPM, at 65 deg F the service water temperature and flow would be sufficient to maintain the motor oil bath temperature below 200 deg F. During the event, the actual event parameters (cooling water flow rate >1 gpm and temperatures< 65 deg F) were less severe than the test parameters and therefore are bounded by the test. Since there was no impact on the RHRSW system's ability to provide cooling water to the RHR system, the RHR system maintained the ability to provide shutdown cooling and residual heat removal. Therefore the event can be retracted since the condition that was reported in the initial event notification report would not have resulted in the prevention of the fulfillment of a safety function (residual heat removal). The licensee notified the NRC Resident Inspector and will notify the State of Minnesota. Notified R3DO(Lara) | |
| ENS 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. | |
| ENS 45247 | 6 August 2009 11:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Maintenance to Affect Technical Support Center | On 8/6/09 at 0600 hours CDT the Monticello Nuclear Generating Plant's Technical Support Center uninterruptable power supply will be isolated to perform a planned maintenance activity. The maintenance activity requires implementation of compensatory measures to maintain TSC functions during the activity. The compensatory measures include having the Emergency Director report to the Control Room and co-locating the remaining TSC staff at the EOF should an event be declared requiring ERO activation. The ERO has previously successfully demonstrated the ability to implement these compensatory measures. The maintenance activity is scheduled to be completed with the TSC returned to full functionality by the end of the dayshift on 8/6/09. The Site Emergency Response Organization has been notified of the maintenance activity and instructed on the planned compensatory measures to be implemented during the activity. MNGP will notify the NRC upon completion of the activity restoring full TSC operability. This event is considered reportable per 10CFR50.72(b)(3)(iii). The licensee has notified the NRC Resident Inspector. * * * UPDATE FROM D. BARNETT TO P. SNYDER AT 1943 ON 8/6/09 * * * On 8/6/09 at 1700 CDT the Monticello Nuclear Generating Plant's Technical Support Center was returned to service and declared functional. The Monticello Emergency Response Organization has been notified and the compensatory measures that were in effect have been terminated. The licensee has notified the NRC Resident Inspector. Notified R3DO (Skokowski). | |
| 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 45338 | 10 September 2009 14:13:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | State Offsite Notification Due to Not Meeting Permit Requirements | The Minnesota State Pollution Control Agency and State Department of Emergency Management were notified today, September 10, 2009, that Monticello did not meet the National Pollutant Discharge Elimination System (NPDES) Permit. Samples from a new monitoring well near the reactor building showed low levels of tritium greater than normal background but below Environmental Protection Agency drinking water standards. The concentration of tritium is below any radiological reporting levels established in station procedures. No elevated levels have been detected in any of the other permanent plant monitoring wells. Therefore, we have no indications that there has been a release of tritium beyond the site from this source. The station will continue to monitor, sample and investigate the source of the tritium. The licensee notified the NRC Resident Inspector.
A subsequent water sample from a new monitoring well near the reactor building indicated a level of 21,300 picocuries/liter of tritium which is slightly above the Environmental Protection Agency drinking water standard for tritium of 20,000 picocuries/liter. No elevated levels have been detected in any of the other permanent plant monitoring wells. Therefore, (the licensee has) no indication that there has been a release of tritium beyond the site from this source. The station will continue to monitor, sample and investigate the source of the tritium. This poses no immediate safety concern for plant employees or the general public. The licensee notified the NRC Resident Inspector. Notified the R3DO (Ring). | |
| ENS 45420 | 7 October 2009 21:41:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Report of Failure of Random Drug Test Non-Licensed Contractor | A non-licensed contractor had confirmed positive for illegal drugs a random fitness for duty test. The contractors access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. |