Semantic search
| Entered date | Site | Region | Reactor type | Event description | Topic | |
|---|---|---|---|---|---|---|
| ENS 39970 | 1 July 2003 05:04:00 | Monticello | NRC Region 3 | GE-3 | 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. | Unanalyzed Condition |
| ENS 40008 | 22 July 2003 21:10:00 | Monticello | NRC Region 3 | GE-3 | A High Energy Line Break (HELB) door was not latched as required. This condition is being reported as an event that could have prevented the fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(A). The door is currently closed. The HELB door which separates two critical Motor Control Center (MCC) areas was unlatched for less than two (2) minutes. The licensee will inform the state representative and has informed the NRC resident inspector.
Because plant safety was not significantly degraded, this event is not reportable under the unanalyzed condition criteria based on: (1) the door in either event was in an uncontrolled condition for less than one minute, (2) the door was not materially affected, only operated improperly, (3) the PRA significance of the event was low, and (4) the HELB Barrier door was not open for a period longer than is allowed by station procedural guidance. R3DO (C. Miller) notified. The station continues to review the event in the station's corrective action program. The NRC Resident Inspector was notified of this retraction by the licensee. | Unanalyzed Condition |
| ENS 40145 | 11 September 2003 00:32:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | Vital Area unescorted access inappropriately granted. Immediate compensatory measures taken upon discovery. The licensee informed the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details. | |
| ENS 40146 | 10 September 2003 21:02:00 | Monticello | NRC Region 3 | GE-3 | Potential break of Fire Water Main in Admin Bldg could cause both divisions of Safe Shutdown equipment to be inoperable by flooding the battery rooms for #11 & 12 125VDC Batteries. Vulnerable section of piping has been isolated to eliminate the flooding concern. Fire Protection compensatory actions have been established in accordance with the site Fire Protection Program. The licensee informed the NRC resident inspector.
Monticello is retracting the event reported based on evaluations which indicate that the fire main is not considered a potential flooding source. Additional evaluations are ongoing, and issues will be entered into the station's corrective action program. The licensee has notified the NRC Resident Inspector. Notified R3DO (Hills) | Safe Shutdown Unanalyzed Condition Fire Protection Program |
| ENS 40481 | 26 January 2004 20:20:00 | Monticello | NRC Region 3 | GE-3 | 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 11:49:00 | Monticello | NRC Region 3 | GE-3 | 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 21:22:00 | Monticello | NRC Region 3 | GE-3 | 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 11:11:00 | Monticello | NRC Region 3 | GE-3 | 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 16:57:00 | Monticello | NRC Region 3 | GE-3 | 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. | Fitness for Duty |
| ENS 40928 | 6 August 2004 11:08:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | This is a notification made under 10CFR 50.72(b)(2)(xi) as a media sensitive event. At 0817 an inadvertent telephone communication to site ERO responders was made. The telephone notification erroneously reported that a security event had occurred at the Prairie Island Nuclear Plant, that the ERO was being activated, and that responders were to report to the back up EOF. No security event has occurred at the Prairie Island Nuclear Plant. Local media has been notified regarding the inadvertent notification and that no security event has occurred at the Prairie Island Nuclear Plant. No news release is planned. This erroneous notification occurred during a training exercise for Security Emergency Communicators. The notification that went out to ERO responders did not identify the message as a drill message. Licensee notified the NRC Resident Inspector, State and Local Emergency Response organizations and local governments. | |
| ENS 41005 | 1 September 2004 17:12:00 | Monticello | NRC Region 3 | GE-3 | 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. | Safe Shutdown Unanalyzed Condition Fire Barrier Fire Protection Program Hourly Fire Watch |
| ENS 41062 | 22 September 2004 13:09:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was obtained from the licensee via facsimile: At approximately 0900 CDT on September 22, 2004, the Goodhue County sheriff's office inadvertently activated the Goodhue County sirens during the conduct of their weekly silent siren test. As a result of the inadvertent siren actuation, Prairie Island Nuclear Generating Plant communicated with the Goodhue County Administrator and the City of Red Wing Mayor's office, as well as a Prairie Island Indian Community representative. No press release is planned. The NRC senior resident inspector was notified. | Siren |
| ENS 41127 | 16 October 2004 00:57:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | At 2215 (CDT) on 10/15/04, the Prairie Island Control Room was notified by the Pierce County Sheriff's Department of a spurious actuation of one EP siren. The siren is located near Highway 63 and County KK. The siren was de-activated by the Pierce County Sheriff personnel. There is no event in progress and no risk to the health and safety of the public. Members of the public notified the sheriff's department of the malfunctioning siren. The siren remains de-activated, is the only non-functional siren, and is expected to be repaired by Nelson Communication on 10/16/04. No further actions are required. The licensee notified the NRC Resident Inspector.
At 0100 on 10/16/04, the same EP siren spuriously re-actuated. The siren was again de-activated by Pierce County Sheriff's personnel. The repair company, Nelson Communication, also dispatched an individual to verify the siren was properly deactivated to preclude additional spurious actuations. The licensee will notify the NRC Resident Inspector. Notified R3DO ( R. Lanksbury) | Siren |
| ENS 41204 | 17 November 2004 23:46:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | Initiation of a Unit 2 shutdown commenced at 2050 on November 17, 2004, as required by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 because both trains of containment cooling were declared inoperable as of 1615. Each train of Prairie Island's Containment Cooling System includes two containment fan coil units (CFCUs), for a total of four CFCUs. TS 3.6.5 requires two trains of containment cooling to be operable. Leakage from the 23 CFCU was determined to render one train of containment cooling inoperable at 1551. At 1615, the 22 CFCU (one of two CFCUs in the opposite train of containment cooling) was also determined to be leaking rendering the second train of containment cooling inoperable. Two CFCUs are fully capable of removing the design basis heat load; however, as the two remaining operable CFCUs are in different trains, neither train of containment cooling could be considered operable. Both inoperable CFCUs have been isolated to restore containment integrity and facilitate repairs. TS LCO 3.6.5 does not have a Condition for both containment cooling trains inoperable, therefore TS LCO 3.0.3 was entered. The licensee has notified the NRC Resident Inspector. | |
| ENS 41267 | 15 December 2004 14:00:00 | Monticello | NRC Region 3 | GE-3 | 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. | Time of Discovery |
| ENS 41374 | 4 February 2005 17:25:00 | Monticello | NRC Region 3 | GE-3 | 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). | Unanalyzed Condition |
| ENS 41377 | 6 February 2005 00:43:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | This report is being made pursuant to 10CFR50.72(b)(3)(ii)(B). At 20:46 on 02/05/05 while reviewing Operation Experience from another Nuclear Site, it was discovered that the design of the AC Auxiliary Power System incorporates a common circuit which could result in a bus lockout preventing the re-energization of both unit one safeguards buses from either onsite or offsite power sources due to a single failure in the common portion of the circuitry. This common circuit includes various metering circuits and is also connected to overcurrent devices that feed breaker lockout relays. Should a failure occur on these common circuits, all breakers supplying power to these buses would be opened, locked out, and prevented from reclosure onto the buses. Due to the loss of the ability to accommodate a single failure, one offsite power source ( CT11 ) has been declared inoperable and Technical Specification 3.8.1 Condition A, entered. Corrective actions are in progress to isolate the common circuit and eliminate the single point vulnerability. The licensee notified the NRC Resident Inspector. See similar events #41362 (Crystal River), #41366 (LaSalle), #41369 (Quad Cities) ,#41370 (Dresden) and #41377 (Monticello). * * * RETRACTION FROM S. SEILHYMER TO S. ROTTON AT 1140 ON 6/13/06 * * * The event reported on February 6, 2005 (4160 volt relaying and metering single failure vulnerability, NRC Event Number 41377) is hereby retracted. Subsequent analysis has concluded that a common mode failure resulting in a simultaneous lockout of both safeguards 4160 volt Bus 15 and Bus 16 due to the common metering circuit was not a credible event in the as-found condition. The analysis showed that the required magnitude of a fire induced fault is much greater than the available fault current in any of the associated circuits. The analysis also showed that dual bus lockout due to an open circuit under non-fire induced conditions is not credible since the imbalance current will be less than the trip setpoint of the affected relays. Thus, although the plant had been in an unanalyzed condition, that condition has now been shown to have not significantly degraded plant safety and is, therefore, not reportable under the requirements of 10 CFR 50.72(b)(3)(ii)(B). Additionally, NMC will be submitting a letter to cancel the associated LER 1-05-01. The licensee notified the NRC Resident Inspector. Notified R3DO (Louden). | Unanalyzed Condition |
| ENS 41436 | 23 February 2005 19:20:00 | Monticello | NRC Region 3 | GE-3 | 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. | Safe Shutdown Unanalyzed Condition |
| ENS 41441 | 24 February 2005 22:39:00 | Monticello | NRC Region 3 | GE-3 | 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. | |
| ENS 41468 | 8 March 2005 11:41:00 | Monticello | NRC Region 3 | GE-3 | 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 41548 | 30 March 2005 09:35:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was obtained from the licensee via facsimile (licensee text in quotes): The site determined that previous repairs to 21, 22, and 23 Containment Fan Coil Unit (CFCU) Heat Exchangers were not in compliance with applicable ASME codes. The three CFCUs were declared inoperable per Generic Letter 90-05 for ASME Class II Piping. Shutdown of Unit 2 to Mode 3 commenced at 0735. Code repairs are being performed and the technical specification required shutdown (LCO 3.0.3/3.6.1.B) is expected to be stopped once one CFCU train is returned to operability and the other is isolated to meet technical specification LCO 3.6.1. 'Containment.' The licensee notified the NRC Resident Inspector. | |
| ENS 41567 | 5 April 2005 22:26:00 | Monticello | NRC Region 3 | GE-3 | 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. | Safe Shutdown Unanalyzed Condition |
| ENS 41606 | 15 April 2005 22:33:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | Unit 2 Train A emergency diesel generator D5 was removed from service at 0838 CDT on 4/11/05 for surveillance testing and Technical Specification (TS) 3.8.1, 'AC Source - Operating,' Condition B, 'One DG inoperable,' was entered. TS Required Action 3.8.1.B.4 requires D5 be restored to operable status with a Completion Time of 7 days. At 1026 CDT, the test was halted due to high-indicated crankcase pressure on Engine 2 (D5 is a tandem engine generator). The test procedure specifies shutting down the DG if crankcase pressure exceeds 30 mm for more than a few minutes (the setpoint for the crankcase pressure trip is 52 mm). Investigation of the cause of the high-indicated crankcase pressure on Engine 2 (and whether Engine 1 was effected) started immediately. Unit 2 Train B emergency diesel generator was demonstrated operable by completing a surveillance test at 0423 CDT on 4/12/05. Evaluation of the scope of work to return D5 to operable status and the schedule for completing the work indicated that repairs could not be completed within the 7 days allowed outage time. Based on this assessment an orderly shutdown of Unit 2 is being performed. Shutdown of Unit 2 commenced at 21:30 CDT on 4/15/05. Unit 2 shutdown will continue until D5 is restored to operable status. The licensee intends to issue a press release. The licensee notified the NRC Resident Inspector. | |
| ENS 41897 | 5 August 2005 22:46:00 | Monticello | NRC Region 3 | GE-3 | 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 19:49:00 | Monticello | NRC Region 3 | GE-3 | During an inspection of structural steel in the Emergency Filtration Train (EFT) building, it was determined that a portion of the steel did not have adequate fire retardant material protecting the steel. This condition would have compromised the 3 hour-fire barrier between the 2 divisions of the Emergency Service Water (ESW) System, impacting # 11 Emergency Diesel Generator ESW Pumps, # 13 and # 14 ESW Pumps. This condition could have resulted in the inability to establish and maintain cold shutdown conditions in the event of a fire in this area. A fire impairment was entered and an hourly fire watch was established to address this condition. All ESW divisions are presently operable. The NRC Resident Inspector will be notified of this event by the licensee.
Monticello is retracting the event notification based on further investigation of the issue. The station has completed calculations that confirm the affected steel beams are not required to maintain fire barrier integrity. Further, the fire severity in the zone will not cause the steel beams to fail. Therefore, the ability of the station to establish and maintain cold shutdown conditions in the event of a fire in this area was not impacted. Based on this information, Monticello has determined there was no unanalyzed condition as reported in Event Notification # 42013. The degraded fire retardant material has been replaced. The licensee notified the NRC Resident Inspector. | Unanalyzed Condition Fire Barrier Fire impairment Hourly Fire Watch |
| ENS 42188 | 5 December 2005 10:45:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The Plant Process Computer (PPCS) will be taken out of service for an approximately 2 week period to implement a planned modification. The current PPCS is being replaced and the computer outage is required to allow cutover to the new PPCS. During this time period ERDS and SPDS will not be available. ERDS and SPDS parameters will be monitored by control board indications. Compensatory actions have been developed. This is an 8-hour reportable event per 10 CFR50.72(b)(3)(xiii) Major Loss of Assessment Capability. The operation of plant systems will not be affected due to this planned action. The PPCS outage is expected to commence around 0600 CST on 12/5/05. The licensee has informed the NRC Resident Inspectors of the modification and schedule. The NRC was previously notified of this planned outage via letter dated November 7, 2005. The NRC Resident Inspector was notified of this event by the licensee.
The Plant Process Computer System was restored at 2330 CST on 12/15/05. The Safety Parameter Display System (SPDS) has been returned to service. The Emergency Response Data System (ERDS) portion of the system will be turned over once the testing has been completed. The licensee will notify the NRC Resident Inspector. Notified R3DO (Burgess).
This notification is an update to previously made notification, EN# 42188. As of 14:30 CST on 1/6/06 the Unit 1 Emergency Response Data System (ERDS) has been accepted for use and is considered fully operational. The licensee notified the NRC Resident Inspector. Notified the R3DO (Ring) | |
| ENS 42250 | 11 January 2006 14:16:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The Unit 2 Plant Process Computer System (PPCS) will be taken out of service for an approximate 2 week period to implement a planned modification. The current PPCS is being replaced and the computer outage is required to allow cutover to the new PPCS. During this time period ERDS and SPDS will not be available. Unit 2 ERDS and SPDS parameters will be monitored by control board indications. Compensatory actions have been developed. This is an 8-hour reportable event per 10 CFR50.72(b)(3)(xiii) Major Loss of Assessment Capability. The operation of plant systems will not be affected due to this planned action. The Unit 2 PPCS outage started at 0743 CST on 1/11/2006. The licensee has informed the NRC Resident Inspectors of the modification and schedule. The NRC was previously notified of this planned outage via letter dated November 7, 2005.
The Unit 2 Plant Process Computer System (PPCS) was restored at 1801 CST on 01/18/06. The Unit 2 Safety Parameter Display System (SPDS) and Emergency Response Data System (ERDS) have also been returned to service. The licensee notified the NRC Resident Inspector. Notified R3DO (M Phillips) | |
| ENS 42301 | 2 February 2006 04:51:00 | Monticello | NRC Region 3 | GE-3 | 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 42312 | 5 February 2006 14:59:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | Unit 2 Train B emergency diesel generator D6 was removed from service at 2010 CST on 1/29/06 for planned maintenance and Technical Specification (TS) 3.8.1, 'AC Source - Operating,' Condition B, 'One DG inoperable,' was entered. TS Required Action 3.8.1.B.4 requires D6 be restored to operable status with a Completion Time of 7 days. The planned maintenance activities included replacing two sets of two pistons, rings and cylinder liners on Engine 2 of D6 (D6 is a tandem-engine diesel generator). Return-to-service testing was initiated on 2/4/06 and at approximately 0000 CST; the test was halted due to high-indicated crankcase pressure on Engine 1. The test procedure specifies shutting down the diesel generator if crankcase pressure on either engine exceeds 30mm for more than a few minutes (the setpoint for the crankcase pressure trip is 52 mm). Investigation of the cause of the high-indicated crankcase pressure on Engine 1 started immediately. Unit 2 Train A emergency diesel generator (D5) was demonstrated operable by completing a surveillance test at 1507 CST on 2/4/06. Evaluation of the scope of work to return D6 to operable status and the schedule for completing the work indicated that repairs could not be completed within the remainder of the 7-day Completion Time. Based on this assessment an orderly shutdown of Unit 2 is being performed. Shutdown (to mode 5) of Unit 2 commenced at 1336 CST on 2/5/06. Unit 2 shutdown will continue until D6 is restored to operable status. The licensee notified the NRC Resident Inspector. | |
| ENS 42504 | 14 April 2006 16:07:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | At 1425 on April 14, 2006, a lockout trip of 11 Condensate Pump occurred. The condensate pump trip caused an expected lockout trip of 11 Main Feedwater Pump trip. With the loss of 50% of feedwater pump capacity, the Shift Supervisor directed a manual Unit 1 reactor trip. The manual reactor trip was successful and all systems responded as expected. The reactor protection system actuation is reportable under 10CFR 50.72(b)(2). A reactor trip from full power results in an expected steam generator narrow range level shrink to 0%. This resultant narrow range steam generator level caused an expected Auxiliary Feedwater System Actuation. Both 11 and 12 Auxiliary Feedwater Pumps started as expected. Auxiliary feedwater actuation is reportable under10CFR 50.72(b)(3). Investigation is underway to determine the cause of 11 Condensate Pump lockout. Plant operations are underway per emergency procedure 1ES-0.1, Reactor Trip Recovery, and 1C1.3, Unit 1 Shutdown, to stabilize the plant in Mode 3, Hot Standby. All control rods fully inserted. Steam generators are discharging steam to the condenser steam dump system. The Auxiliary Feedwater Pumps are maintaining Steam Generator level. The electrical grid is stable. The licensee will notify the NRC Resident Inspector. | |
| ENS 42565 | 10 May 2006 16:54:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | This notification is being made in accordance with 10CFR50.72(b)2(xi) due to a press release being issued. The Nuclear Management Company (NMC) will be issuing a press release regarding maintenance activities on a non-safety related, non-radioactive system that will state: 'During work on a plant equipment modification at Prairie Island Nuclear Plant Unit 1 last week, some gasket material being removed was found to contain asbestos. Prairie Island Unit 1 is in the midst of a refueling outage. 'As soon as site management became aware of the concern for potential asbestos containing material, work on the project was suspended, according to Nuclear Management Co., which operates the plant. The air immediately was monitored and tested according to Minnesota Clean Air requirements. The results were well below regulatory limits. However, results from additional testing on surfaces by a certified lab indicated detectable levels of asbestos on some surfaces in the immediate area. 'Trained asbestos workers are performing a complete clean-up of the affected area. 'NMC and its contractor are taking precautionary actions for any employee with a concern for potential exposure. This includes offering an optional asbestos exposure examination in accordance with federal Occupational Safety and Health Administration (OSHA) guidelines. 'Prairie Island's two reactors near Red Wing, Minn., generate 1,076 megawatts of electricity, enough to power one in five homes and businesses in the Upper Midwest. Xcel Energy owns the plant, which is one of six reactors operated by Nuclear Management Company, headquartered in Hudson, Wis.' Unit 1 is currently defueled. The licensee informed state/local agencies/nearby tribal entity and the NRC Resident Inspector. | |
| ENS 42607 | 28 May 2006 01:10:00 | Monticello | NRC Region 3 | GE-3 | 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 04:06:00 | Monticello | NRC Region 3 | GE-3 | 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 | 31 July 2006 00:11:00 | Monticello | NRC Region 3 | GE-3 | 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 42754 | 7 August 2006 21:20:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | While performing routine annual maintenance on the plant heating boiler, water that is normally discharged to the turbine building sump was redirected through a drain to gravel just outside the turbine building. This water contained low levels of tritium. This is a voluntary notification even though the tritium levels were below the EPA guidelines for drinking water and no permits were violated. No follow up action is necessary and monitoring will continue. On July 31 (2006) the site joined the industry initiative regarding lower thresholds for reporting tritium and the ground water reassessments being completed at all nuclear plants in the industry. Tritium level in the water was 1.91E-5 microCuries/ml. The licensee notified the NRC Resident Inspector. | |
| ENS 42788 | 19 August 2006 23:10:00 | Monticello | NRC Region 3 | GE-3 | 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 43010 | 25 November 2006 04:14:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | With Prairie Island Unit 2 core offloaded for the 2R24 refueling outage, it was discovered that a potential common cause failure of the 21 and 22 Residual Heat Removal (RHR) Pumps may exist. A non-safety related 120 VAC motor heater circuit from Panel 2RPA3 Circuit 28 supplies power to 21 and 22 RHR Pumps. Investigation continues. The RHR system is not required by the current operating mode. The RHR Pumps on the operating unit (Unit 1) are supplied from separate 120 VAC panels so the concern does not exist for the 11 and 12 RHR pumps. The licensee notified the NRC Resident Inspector.
Event Number 43010 (reported 11/24/06) reported a potential single failure vulnerability of the Unit 2 residual heat removal (RHR) pumps. NMC identified that power cables for the motor heaters in each pump were powered from the same 120 Volt panel and that the heater power cables were routed for some distance in the same conduit. The concern was that an electrical fault could propagate from one RHR pump motor to the opposite train RHR pump motor. Further engineering evaluation has determined that no credible failure mechanism exists that could cause a fault in one RHR pump motor to propagate to the opposite train RHR pump motor. Since the dielectric strength of insulation on the heaters is greater than the dielectric strength of the air between the motor windings and the motor frame or the stator, any fault in the motor windings would be to ground (and, thus, interrupted by the ground fault protective relaying). Therefore, no potential single failure vulnerability existed in the as-found condition and NMC hereby retracts the event reported on 11/24/06 (Event Number 43010). The licensee notified the NRC Resident Inspector. Notified the R3DO (H. Peterson). | Unanalyzed Condition |
| ENS 43088 | 10 January 2007 19:46:00 | Monticello | NRC Region 3 | GE-3 | 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 19:09:00 | Monticello | NRC Region 3 | GE-3 | 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 14:27:00 | Monticello | NRC Region 3 | GE-3 | 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 12:35:00 | Monticello | NRC Region 3 | GE-3 | 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 04:41:00 | Monticello | NRC Region 3 | GE-3 | 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 43253 | 21 March 2007 20:05:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The Prairie Island Nuclear Plant has a turbine building sump that collects water from the secondary side of the plant and this water contains trace levels of tritium. When we normally discharge the water from the sump, it is either sent to the river or to a permitted area on plant property. Earlier today, the discharge was lined up to the permitted area on plant property. The end of the line was frozen, causing approximately 100 - 150 gallons of the water to be diverted to a non-permitted area inside the plant property. This notification is being made due to the notification of offsite agencies (state , local and county). The NRC Resident Inspector was notified of this event by the licensee. | |
| ENS 43280 | 5 April 2007 14:01:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | At 09:08 am on 4/5/2007, during surveillance testing of Unit 2 Train A safeguards logic at power, a spurious Train A safety Injection (SI) actuation occurred resulting in reactor protection system (RPS) actuation. Train A SI was in "Test" at the time and should not have caused the RPS trip. The operating crew manually actuated Train B SI as required by emergency operating procedures. All automatic actions for a reactor trip and safety Injection occurred as required. Reactor Coolant System (RCS) pressure decreased below the shutoff head of the high head Emergency Core Cooling System (ECCS) pumps during the transient, resulting in momentary ECCS discharge to the RCS. SI has been terminated per emergency operating procedures. Prairie Island Unit 2 has been stabilized in mode 3, at about 2235 psig and 547 degrees average RCS temperature. Decay heat Is currently being removed by auxiliary feedwater and secondary steam dump to the main condenser. The cause of the actuation signal is under investigation. All control rods fully inserted. No primary power operated relief valves or safety valves lifted. No steam generator safeties lifted. Safeguards buses are powered by offsite power. The Unit 2 Emergency Diesel Generators (EDG) started but did not load. Unit 1 Control Rod Drive Mechanism cooling isolated as designed in response to the actuation and has since been restored. Otherwise, Unit 1 was unaffected and remains in mode 1 at 100% power. The licensee notified the NRC Resident Inspector. The licensee will also be notifying the State, local and other Government agencies and will be issuing a press release. | |
| ENS 43404 | 4 June 2007 17:28:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | During performance of NFPA-805 Transition Project Task SUP-1, 'Manual Action Compliance,' it was determined there were 5 non-compliant manual operator actions that were being performed to achieve and maintain hot safe shutdown in Fire Area 29, Admin Building Electrical & Piping Room 1. These manual actions were being performed 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.' The discovery of these non-compliant manual actions is being reported as an unanalyzed condition as defined by 10 CFR 50.72(b)(3)(ii). The alternate compensatory measure for Fire Area 29 is to perform the manual action. An extent of condition review has been initiated that will encompass the remainder of the safe shutdown areas in both Prairie Island Nuclear Generating Plant (PINGP) units. 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. The licensee will notify the NRC Resident Inspector. * * * UPDATE FROM W. BODIN TO P. SNYDER AT 1643 ON 6/5/07 * * * During performance of NFPA-805 Transition Project Task SUP-1, 'Manual Action Compliance,' it was determined that non compliant manual operator actions are credited to achieve and maintain hot safe shutdown for a fire. The following actions were identified during the extent of condition reviews conducted subsequent to EN 43404 report. Numbers of Non-compliant Actions and Fire Areas Affected: 2 actions for Fire Area 2, Ventilation Fan Room, Unit 1; 2 actions for Fire Area 10, Train A Event Monitoring Equipment Room; 6 actions for Fire Area 20, Unit 1 4.16 KV Safeguards Switchgear (Bus 16); 2 actions for Fire Area 22, 480V Safeguards Switchgear (Bus 121); 1 action for Fire Area 23, Unit 2, 4.16 KV Normal Switchgear (Bus 23, 24); 2 actions for Fire Area 30, Administration Building Electrical & Piping Room #2; 1 action for Fire Area 33, Battery Room 11; 1 action for Fire Area 34, Battery Room 12; 6 actions for Fire Area 37, Unit 1 480V Normal Switchgear Room; 4 actions for Fire Area 41A, Screenhouse (DDCWP Rooms); 6 actions for Fire Area 41B, Screenhouse Basement; 10 actions for Fire Area 58 and 73, (Auxiliary) Building Ground Floor Unit 1 & 2; 2 actions for Fire Area 59, (Auxiliary) Building Mezzanine Level Unit 1; 1 action for Fire Area 80, 480V Safeguard Switchgear Room (Bus 112); and 4 actions for Fire Area 81, 4.16 KV Safeguard Switchgear Room (Bus 15). These manual actions are credited for safe shutdown 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.' The discovery or these non-compliant manual actions is being reported as an unanalyzed condition as defined by 10 CFR 50.72(b)(3)(ii). The manual actions have been determined to be feasible and will be treated as alternate compensatory measures for Fire Areas 2, 10, 20, 22, 23, 30, 33, 34, 37, 41A, 41B, 58, 59, 73, 80, and 81. An extent of condition review is continuing that will encompass the remainder of the safe shutdown areas in both Prairie Island Nuclear Generating Plant (PINGP) Units. The licensee notified the NRC Resident Inspector. Notified R3DO (M. Phillips).
During performance of NFPA-805 Transition Project Task SUP-1, 'Manual Action Compliance,' it was determined that noncompliant manual operator actions are credited to achieve and maintain hot safe shutdown for a fire. The following actions were identified during the extent of condition reviews conducted subsequent to EN 43404 report.
1 action for Fire Area 4, Fuel Handling Area 12 actions for Fire Area 31, 'A' Train Hot SD panel, AFW & Air Compressors 11 actions for Fire Area 32, 'B' Train Hot SD panel, AFW & Air Compressors 2 actions far Fire Area 38, Unit 2 480V Normal Switchgear Room 1 action for Fire Area 39, Radwaste Building 2 actions for Fire Area 46, Cooling Tower Equipment House 4 actions for Fire Area 69, Turbine Building Ground & Mezzanine Floors Unit 1 3 actions for Fire Area 70, Turbine Building Ground & Mezzanine Floors Unit 2 2 actions for Fire Area 71, Containment Unit 2 2 actions for Fire Area 72, Containment Annulus Unit 2 5 actions for Fire Area 74, Aux Building Mezzanine Level Unit 2 These manual actions are credited for safe shutdown 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.' The discovery of these non-compliant manual actions is being reported as an unanalyzed condition as defined by 10 CFR 50.72(b)(3)(ii). The manual actions have been determined to be feasible and will be treated as alternate compensatory measures for Fire Areas 4, 31, 32, 38, 39, 46, 69, 70, 71, 72, and 74. This concludes the extent of condition review of the safe shutdown areas in both Prairie Island Nuclear Generating Plant (PINGP) units. The licensee notified the NRC Resident Inspector. Notified the R3DO (Phillips). | Safe Shutdown Unanalyzed Condition Operator Manual Action Manual Operator Action |
| ENS 43488 | 12 July 2007 19:14:00 | Monticello | NRC Region 3 | GE-3 | 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. | Safe Shutdown Unanalyzed Condition Operator Manual Action Manual Operator Action |
| ENS 43492 | 16 July 2007 18:09:00 | Monticello | NRC Region 3 | GE-3 | 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 43500 | 17 July 2007 19:10:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | A small fire was discovered in an office building outside the protected area. The fire was extinguished using a hand held water fire extinguisher. The office building was evacuated as a precaution. One individual received a minor burn and was tended to by an onsite EMT. Local fire department responded to assist in smoke removal. EAL classification did not apply since the fire was outside the protected area. Fire was reported to be a trash can fire with fire damage being limited to the trash can, adjacent wall, and ceiling tile. Courtesy notifications to state and local government agencies have been made. The licensee notified the NRC Resident Inspector. | |
| ENS 43515 | 24 July 2007 10:40:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | During performance of the Technical Support Center (TSC) Ventilation System Operability Test, an outside air damper failed to close causing a failure of the TSC Ventilation System to attain the required 0.125 inches water column positive pressure. If an emergency condition occurs during the time the repairs are being made, plans are to utilize the TSC as long as radiological conditions allow. Procedure F3-6 ACTIVATION AND OPERATION OF THE TSC, section 7.6, directs TSC management to relocate TSC activities to a radiological safe area if necessary. The licensee notified the NRC Resident Inspector.
Technical Support Center (TSC) Ventilation System is now operable. R3DO (Steve Orth) notified. The NRC Resident Inspector was notified of this event update by the licensee. | |
| ENS 43525 | 26 July 2007 16:32:00 | Monticello | NRC Region 3 | GE-3 | 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. | Unanalyzed Condition |