|Report date||Site||Event description|
|05000306/LER-2017-002||11 December 2017||Prairie Island|
On October 16, 2017, with Unit 2 shutdown for a refueling outage, investigation into a boric acid indication identified a through wall leak at the socket weld that joins a 3/4 inch line to Loop A Reactor Coolant System (RCS)(AB) shutdown communication line valve 2RC-8-37 )(VTV). The leak was isolated by closed valves that would have limited primary coolant leakage to within the capacity of the charging system when the reactor coolant system was pressurized. The quantity of dry boric acid at the location was small (estimated at 1/2 teaspoon in volume). This failure constituted a welding or material defect in the primary coolant system that was not found acceptable under ASME Section Xl and an event or condition prohibited by Technical Specifications.
The cause of the leakage was determined to be stress corrosion cracking. Valve 2RC-8-37 was replaced. In addition, Prairie Island Nuclear Generating Plant intends to perform phased array ultrasonic inspections of socket welds on similar Class 1 piping containing stagnant water during future refueling outages.
|05000306/LER-2017-001||29 November 2017||Prairie Island|
From May 2, 2016 to May 6, 2016, when B Train 122 Control Room Chiller (CRC) was out-of-service (OOS) per Technical Specifications (Tech Specs) 3.7.11 Condition A, Unit 2 A Train 23 Containment Fan Coil Unit (FCU) was OOS. According to revision 41 of site procedure C18.1, "Engineered Safeguards Equipment Support Systems," Bus 16 load sequencer and Bus 121 were inoperable when 122 CRC was OOS. Bus 121 supports B Train Diesel Driven Cooling Water Pump and Unit 2 B Train containment cooling (22/24 FCUs). So both trains of containment FCUs were OOS at the same time for approximately 35.6 hours. This would have required entry into LCO 3.0.3 putting Unit 2 in MODE 3 within 7 hours, this did not occur. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Operation or Condition Prohibited by Tech Specs.
The cause was that the Senior Reactor operators failed to utilize Human Performance Tools (Verification/Validation and Procedure Use/Adherence) when assessing the Tech Specs impact to Unit 2 for applying LCO 3.0.6 when 122 CRC was taken OOS.
Corrective actions include independent assessment of shared system LCO's for each unit, revising the LCO database, established a standard for LCO 3.0.6 log entries, and revising the safety function determination program to be more user friendly.
|05000282/LER-2016-006||15 February 2017||Prairie Island|
On December 18, 2016, at 0818 CDT, the Prairie Island Nuclear Generating Plant (PINGP) had a fire in the PINGP Switch Yard due to 8H8 Breaker-CT (Current Transformer) catastrophic failure. During the event, 121 Motor Driven Cooling Water Pump (MDCLP) auto-started on low header pressure (80 psi) as designed, and supplied load to the Cooling Water (CL) Header to increase header pressure. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of an emergency service water system.
The cause of the 121 MDCLP auto-start was a low-pressure transient in the cooling water pump discharge header that was caused by the trip of the Containment & Auxiliary Building Chiller. When the chiller tripped, Unit 2 Containment Fan Coil units automatically swapped to cooling water. This additional load on the CL system, which caused system pressure to drop below 80 psi. The health and safety of the public was not at risk.
|05000282/LER-2016-005||7 October 2016||Prairie Island|
On August 21, 2016, at 1740 CDT, the Prairie Island Nuclear Generating Plant (PINGP) 2RY Transformer locked out.
During the event, 121 Motor Driven Cooling Water Pump (MDCLP) stopped due to the loss of power and then automatically restarted when sequenced by the load sequencer. The pump auto started on low pressure in the cooling water pump discharge header. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of an emergency service water system.
The cause of the 121 MDCLP auto-start was a low-pressure condition in the cooling water pump discharge header resulting from 2RY Transformer lockout. There was no safety injection signal from PINGP Unit 1 nor Unit 2 when 121 MDCLP started. The health and safety of the public was not at risk. The most likely cause of 2RY Transformer lockout is that the breathers on 2RY Transformer dog house were not adequately maintained. Breathers were not adequately maintained due to design limitations that did not allow proper internal inspection.
Immediate actions taken, 2RY Transformer dog house dried out and internal components were tested to validate the functionality of the equipment. Results were satisfactory. Corrective actions, 2RY Transformer Dog house breathers and outdoor bus breather were replaced per work order prior to returning the transformer to service. Also, the Preventative Maintenance (PM) procedure was updated to perform the full internal inspection during the future PM's (bus duct or transformer).
|05000306/LER-2015-002||22 June 2016||Prairie Island|
On April 3, 2015, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was operating at 100 percent power, when at 0652 CDT, an unexpected annunciator, 47510-0104 21 FEEDWATER PUMP LOCKED OUT was received. The reactor was manually tripped as required by the annunciator response procedure. This also resulted in a turbine trip as designed. The Operations crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3, at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The Auxiliary Feedwater Pumps actuated as designed on low narrow range steam generator level. Steam Generator levels were returned to normal.
This event is reportable under 10 CFR 50.72(b)(2)(iv)(B), any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation, and 10 CFR 50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any of the systems listed in paragraph 10 CFR 50.72(b)(3)(iv)(B)(6), PWR auxiliary or emergency Feedwater system.
The cause evaluation determined that the event was caused by pressure fluctuation within the system which resulted in the bourdon tube movement at a high frequency causing wear of the internal components of the pressure switch.
Corrective actions: Pressure Switch (PS-16012) was replaced immediately on April 3, 2015, and to install snubbers to reduce process flow fluctuations experienced by Feedwater Pump pressure switches.
|05000282/LER-2016-004||21 June 2016||Prairie Island|
On April 21, 2016, during a walkdown of fire barriers for the National Fire Protection Association (NFPA) 805 project, it was determined that the fire barrier between Fire Area (FA) 59 (Auxiliary Building Mezzanine Unit 1) and FA 85 (Holdup Tank/ Demineralizer Area) is not a rated barrier due to unsealed combustible pathway penetrations in the barrier.
The walkdown also identified that penetrations in the fire barriers between FA 68 (Unit 1 Annulus) and FA 60 (Auxiliary Building Operating Level Unit 1), FA 68 and FA 61A (Auxiliary Building Hatch Area), FA 72 (Unit 2 Annulus) and FA 75 (Auxiliary Building Operating Level Unit 2), and FA 72 and 61A are not sealed with fire rated materials.
Both conditions were reported under 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degrades plant safety due to the missing fire barrier between redundant Appendix R safe shutdown trains.
The apparent cause was determined to be that the Engineering Manual 3.4.2 does not require Appendix R program owner review of Fire Protection Engineering Evaluations that depend on Appendix R Safe Shutdown analysis.
The corrective action is to revise the Engineering Manual 3.4.2 to require Appendix R review when the program is impacted.
|05000282/LER-2016-002||25 March 2016||Prairie Island|
At 1110 Central Standard Time (CST) on 1/29/2016, the Prairie Island Nuclear Generating Plant (PINGP) performed a planned overspeed post-maintenance test (PMT) of 22 Diesel-Driven Cooling Water Pump (DDCLP) in accordance with plant maintenance procedure. During the overspeed trip test PMT, 22 DDCLP tripped as expected, and 121 Motor-Driven Cooling Water Pump (MDCLP) unexpectedly started automatically ("auto-started") on low pressure in the cooling water pump discharge header. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of an emergency service water system that does not normally run and that serves as ultimate heat sink.
The cause of the 121 MDCLP auto-start was a low-pressure transient in the cooling water pump discharge when 121 MDCLP started. The health and safety of the public was not at risk.
Corrective actions to be taken for this event include revising plant procedures to ensure 121 MDCLP is running prior to performing the overspeed trip test PMT and to ensure the pump is running or in pullout position to prevent auto-start of 121 MDCLP when stopping a DDCLP.
|05000282/LER-2016-003||4 March 2016||Prairie Island|
On 1/7/2016, Prairie Island Nuclear Generating Plant (PINGP) reviewed corrective actions associated with the transition to National Fire Protection Association "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (NFPA 805). PINGP discovered no cold shutdown repair procedure existed to restore power to train B reactor coolant system (RCS) vent valves to reduce RCS pressure in event of fire in Fire Area (FA) 59 (Auxiliary Building Mezzanine Floor Unit 1) and FA 74 (Auxiliary Building Mezzanine Floor Unit 2). The PINGP Appendix R safe shutdown analysis (SSA) credits such actions for train B RCS vent valves for a fire in these areas. PINGP established compensatory measures for affected equipment and, with automatic detection and suppression capability, these measures ensured protection of potentially affected equipment. The health and safety of the public was not at risk. PINGP submitted Event Notification (EN) 51642 on 1/7/2016 as this event is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. On 04/14/2005, a revision of the SSA was approved that credited a cold shutdown repair 'as a compliance strategy for FA 59 and 74, but no repair procedure existed in the event of afire in FA 59 nor FA 74.
On 1/11/2016, F5 Appendix D "Impact of Fire Outside Control/Relay Room" was issued and includes cold shutdown repair procedures for Train B RCS vent valves. Lack of technical rigor to ensure procedure F5 Appendix D included actions required by the SSA is the identified cause.
|05000282/LER-2016-001||18 February 2016||Prairie Island|
On 12/21/2015 with Unit 1 (U1) in Mode 1 at 100% power and Unit 2 (U2) in Mode 3, a technical review of a new weak link calculation as part of the transition to National Fire Protection Association's Performance Based Standard for Fire Protection for Light Water Reactor Electronic Generating Plants (NFPA 805) identified a non-emergency unanalyzed condition reportable under 10 CFR 50.72(b)(3)(ii)(B). Specifically, several motor operated valves (MOVs) credited to be manually operated from outside the control room in the event of a fire in the control room or relay room could be damaged in a postulated fire if hot shorts were to bypass the torque and limit switches. Also, MOVs associated with the Gland Steam system of both U1 and U2 had been added to F5 Appendix B but never analyzed in the event of a hot short. These conditions could impact the ability of plant operators to implement procedure F5 Appendix B, Control Room Evacuation (Fire). The identified MOVs are located in Fire Area (FA) 13 (Control Room) and FA 18 (Relay and Cable Spreading Room). New hourly fire watch impairments were created in these areas as compensatory measures. The fire detection systems in the control room and relay room remain in service. The site submitted Event Notification (EN) 51616.
Reviews of the list of MOVs susceptible to hot shorts bypassing the torque and limit switches continued and additional valves were noted to be affected by this failure mechanism. As a compensatory measure, new hourly fire watch impairments were created for FA 31 ("A" Train Hot Shutdown Panel & Air Compressor/Auxiliary 695 Feedwater Room), FA 32 ("B" Train Hot Shutdown Panel & Air Compressor/Auxiliary 695 Feedwater Room), FA 58 (Auxiliary Building Ground Floor U1) and FA 73 (Auxiliary Building Ground Floor U2). EN 51616 was updated on 01/14/2016. The public health and safety was not impacted.
During technical review for a new calculation as part of the transition to NFPA-805, the station reviewed MOVs identified in the Appendix R program to ensure they had been analyzed correctly. In this review it was determined several MOVs had been added to the program but had not been analyzed for a weak link. Further, several MOVs would not survive a hot short condition, although they were credited in the Appendix R program to meet programmatic requirements for survivability from a hot short condition. Also, MOVs associated with the Gland Steam system of both U1 and U2 had been added to F5 Appendix B but never analyzed in the event of a hot short. These MOVs could be affected by a fire in FA 13 (Control Room) and FA 18 (Relay and Cable Spreading Room). This unanalyzed condition could impact the ability of plant operators to implement procedure F5 Appendix B. Therefore, the station reported this unanalyzed condition under 10 CFR 50.72(b)(3)(ii)(B) on 12/21/2015 as EN 51616. New hourly fire watch impairments were created in these fire areas as compensatory measures.
After submittal of EN 51616, the review was further expanded for all credited MOVs which resulted in identifying additional MOVs that would not survive a hot short condition if a fire were to occur in the following fire areas: FA 31 ("A" Train Hot Shutdown Panel & Air Compressor/Auxiliary Feedwater Room, 695), FA 32 ("B" Train Hot Shutdown Panel & Air Compressor/ Auxiliary Feedwater Room, 695), FA 58 (Auxiliary Building Ground Floor U1), and FA 73 (Auxiliary Building Ground Floor U2). EN 51616 was updated on 01/14/2016 listing the additional fire areas and the compensatory actions of new hourly fire watch impairments.
On February 28, 1992 the NRC released Information Notice 92-18 (IN 92-18), Potential for Loss of Remote Shutdown Capability During a Control Room Fire, addressing the potential for a control room fire causing electrical short circuits ("hot shorts") between normally energized conductors and conductors associated with the control circuitry of motor-operated valves (MOVs) required for achieving and maintaining post-fire safe shutdown conditions. This could result in a loss of capability to achieve or maintain safe shutdown conditions 2
prompted the 12/15/2015 event notification, and the updated notification on 01/14/2016, for an unanalyzed condition due to non-compliant fire protection manual operator actions reportable under 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures were initiated in affected fire areas.
EN 51616 was submitted to the NRC.
During the causal evaluation for the LER, PINGP reviewed its response to a 1997 Unresolved Issue (URI) regarding the review of MOVs in accordance with IN 92-18 (Accession Number 9710200112). The response stated PINGP would use IN 92-18 as guidance to perform additional reviews of the hot short problem following the completion of the Appendix R safe shutdown analysis and create an accurate list of MOVs susceptible to hot shorts (ADAMS Accession Number 9711200012). On August 7, 1998, calculation ENG-ME-353, Mechanical MOV Analysis to support IN 92-18 determined the valve and operator thrust and torque values that may occur during a hot short that bypasses the limit switches and torque switches. The calculation addressed 46 Appendix R related valves and found 32 could potentially be damaged during a fire induced hot short requiring modifications and revision to the Appendix R safe shutdown analysis.
During a fire, a hot short could cause the MOV to operate without its torque/limit switch protection, damaging the valve and preventing manual operation of the valve. All credited MOVs for safe shut-down were reviewed for this condition and seventeen MOVs were found susceptible.
Susceptible MOVs Eleven (11) MOVs require modification to re-wire the torque and limit switches to prevent damage to the actuator:
3 infocollects.Resource@nrc.goy, and to the Desk Officer, Office of Information and Regulatory Affairs, Prairie Island 05000282 1) MV-32006, 1 TURB GLND STM SL SPLY UPSTRM SO MV 2) MV-32010, 1 TURB GLND STM SL SPLY B-P MV 3) MV-32016, 11 SG MS SPLY TO 11 TD AFW PMP MV 4) MV-32017, 12 SG MS SPLY TO 12 TD AFW PMP MV 5) MV-32019, 21 SG MS SPLY TO22 TD AFW PMP MV 6) MV-32020, 22 SG MS SPLY TO 22 TD AFW PMP MV 7) MV-32021, 2 TURB GLND STM SL SPLY UPSTRM SO MV 8) MV-32022, 2 TURB GLND STM SL SPLY B-P MV 9) MV-32238, 11 AFW TO 11 SG MV 10) MV-32246, 22 AFW TO 21 SG MV 11) MV-32333, 11 TD AFW PMP SUCT FROM CST MV Six (6) MOVs require PINGP to revise the Appendix R safe shutdown analysis to credit actions to isolate other valves or to credit check valves.
1) MV-32115, 122 SFP HX INLT HDR MV B 2) MV-32117, 121 SFP HX INLT HDR MV A 3) MV-32166, 1 REAC EXCS LTDN LINE ISOL MV A 4) MV-32194, 2 REAC EXCS LTDN LINE ISOL MV A 5) MV-32335, 12 MD AFW PMP SUCT FROM CST MV 6) MV-32336, 21 MD AFW PMP SUCT FROM CST MV The causal evaluation for this event determined the cause was the failure to include ENG-ME- 353 as an input to the Appendix R safe shutdown analysis. Because of this exclusion, the site did not update ENG-ME-353 to include new weak link data to support the Appendix R safe shutdown analysis.
Additionally, the causal evaluation identified inadequate supervisory oversight during the transition to NFPA 805 that contributed to the event. When one engineer was assigned to both the Appendix R program and the transition to NFPA 805 resulting in untimely responses, in a 4
Infocollects,Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, 05000282 Prairie Island
manner commensurate with their safety significance, to previously identified deficiencies for compliance with IN 92-18.
The NFPA 805 License Amendment identified the need to revise ENG-ME-353 to incorporate updated vendor information (ADAMS Accession Number ML12278405). During the revision of ENG-ME-353, this issue was identified, There were no structures, components or systems that were inoperable at the start of the event that contributed to the event.
There was no nuclear, environmental, radiological or industrial safety consequences related to this event. PINGP has procedures and controls in places to minimize the likelihood and severity of fires occurring, and a significant fire impacting the ability to safely shutdown did not occur.
This postulated fire scenario requires that a significant fire of sufficient size and intensity to damage MOV control cables and bypass torque and limit switches to cause spurious operation and damage to MOVs. The frequency of fires large enough to damage control cables is much lower than the generic fire ignition frequency that includes fires of all intensities. To damage the MOV, the fire must cause damage to cable insulation to the point that an energized conductor shorts around the torque and limit switches to the MOV control circuit for sufficient time for the MOV to stroke to the opposite position without open circuit or short to ground.
Causes PINGP failed to include ENG-ME-353 as an input to the Appendix R safe shutdown analysis which led to failure to update ENG-ME-353 to include new weak link data.
During the original analysis for IN 92-18, only a specified list of MOVs was reviewed to be credited; however, additional MOVs were added without the required analysis to support survivability.
o Eleven (11) MOVs require modification to re-wire the torque and limit switches to prevent damage to the actuator:
® MV-32006, 1 TURB GLND STM SL SPLY UPSTRM SO MV ® MV-32010, 1 TURB GLND STM SL SPLY B-P MV ® MV-32016, 11 SG MS SPLY TO 11 TD AFW PMP MV ® MV-32017, 12 SG MS SPLY TO 12 TD AFW PMP MV
0 ® MV-32115, 122 SFP HX INLT HDR MV B ® MV-32117, 121 SFP HX INLT HDR MV A ® MV-32166, 1 REAC EXCS LTDN LINE ISOL MV A ® MV-32194, 2 REAC EXCS LTDN LINE ISOL MV A ® MV-32335, 12 MD AFW PMP SUCT FROM CST MV ® MV-32336, 21 MD AFW PMP SUCT FROM CST MV 6
Previous Similar Events
Damage by Fire Induced Hot Shorts" was submitted on September 7, 1998 (ADAMS Accession Number 9809150244).
a. Re-evaluate MOVs to identify alternate shutdown systems, components, or flow paths that are not susceptible to damage and revise the Safe Shutdown Analysis accordingly, or b. Modify MOV's mechanically to prevent mechanical damage (e.g.
smaller motor), or c. Modify MOV's electrically to prevent hot short susceptibility (e.g. hold the MOV circuit breaker open or rewire the MOV control circuit).
submitted on October 26, 1998 (ADAMS Accession Number 9810300149).
7 NRC with a schedule of the corrective actions as committed to in the original submittal (ADAMS Accession Number 9904150006).
|05000306/LER-2016-001||12 February 2016||Prairie Island|
On 12/17/2015 at 1234 Central Standard Time (CST) with Unit 2 (U2) in Mode 1 at 100 percent power, the U2 electric generator tripped due to a ground overvoltage protection relay (59/2G) actuation, which resulted in an automatic reactor trip and expected Auxiliary Feedwater System actuation. At 1307 CST a fire alarm was received in the U2 containment.
Personnel were unable to verify the status within 15 minutes and the site declared an Unusual Event, HU2.1, at 1318 CST.
The licensee notified the NRC Resident Inspector, state and local authorities and made an Event Notification to the NRC for the Unusual Event and the U2 reactor trip as required under 10 CFR 50.72(a)(1)(i) and 10 CFR 50.72(b)(2)(iv)(B).
Subsequently, the Unusual Event was terminated on 12/17/2015 at 1450 CST after a containment entry determined there was no smoke or fire. The original Event Notification was updated to cancel the Unusual Event and to add reporting under 10 CFR 50.72(b)(3)(iv)(A) for the expected Auxiliary Feedwater system actuation. Decay heat removal was via forced circulation with Auxiliary Feed and Steam Dumps providing secondary cooling. Offsite power remained available. The health and safety of the public was not at risk.
A troubleshooting plan was initiated which determined an actual ground fault existed internal to the new U2 electric generator, installed during the Fall 2015 U2 refueling outage, causing a U2 reactor trip and the expected Auxiliary Feedwater actuation. The cause of the ground fault of the U2 electric generator was determined to be foreign material, specifically a hex socket and hex key introduced at the vendor's manufacturing facility during installation of a temporary dust cover on collector end of the generator rotor.
|05000306/LER-2014-002||15 July 2014||Prairie Island||At approximately 1236 CDT on May 19, 2014, a Cooling Water (CL) Leak was identified on the lower face on the northeast corner on 23 Containment Fan Coil Unit (FCU) flange. The FCU is considered part of the Containment boundary. The leak caused a breach to this boundary, which could result in radioactivity traveling from Containment into the CL system, and resulting in off-site dose to the public. Unit 2 Containment was declared inoperable for a loss of Containment integrity, which required entry into Technical Specifications (TS) LCO 3.6.1 Condition A, Containment inoperable, in MODES 1, 2, 3, and 4. Immediate actions were taken to isolate the FCU and it was isolated within 1 hour from the initial conformation of the leak and TS 3.6.1 Condition A was exited. This restored Containment to an operable status. A second leak was found on the southwest CL supply inlet flange. A Work Request was initiated to repair the leak, and 23 Containment FCU was restored to an operable condition. This condition is reportable under 10 CFR 50.73(a)(2)(v)(C), Event or Condition that Could Have Prevented Fulfillment of a Safety Function to control the release of radioactive material. The Root Cause evaluation determined that the maintenance procedures and work plans for installation of Containment FCU Head Face Flange gaskets and inlet/outlet piping spacers/gaskets do not meet station planning standards for critical component maintenance and repair to prevent leakage through all ranges of operation. Planned corrective actions are to revise the Containment FCU Maintenance and Inspection procedures to specify a 45ft-lbs torque value for the Containment FCU Head Face Flanges vice providing an allowable range and to direct workers to take measurements of inlet/outlet flange spacers during installation to ensure spacers are centered in the flange.|
|05000282/LER-2013-001||8 October 2013||Prairie Island|
In December 2010, Control Room Envelope (CRE) testing was conducted to satisfy Surveillance Requirements (SR) 184.108.40.206. At that time, as a prerequisite, the surveillance procedure contained a step to add water to two floor drains loop seals located in the 121 and 122 Safeguards Chiller Room that penetrated the CRE.
On August 9, 2013, this practice was determined to be unacceptable preconditioning. Because the December 2010, surveillance test was the first performance of this surveillance, the unacceptable preconditioning resulted in a never performed surveillance, failure to meet the associated surveillance requirement, and an inoperable CRE.
On September 13, 2013, the CRE was tested and failed due to excessive in-leakage from Control Damper CD- 34177. This condition is reportable per 10 CFR 50.73(a)(2)(v)(C), event or condition that could have prevented fulfillment of a safety function and 10 CFR 50.73(a)(2)(i)(B), operation or condition prohibited by Technical Specification.
Based on engineering's past operability evaluation with the equipment history of CD-34177, the CRE was inoperable from December 10, 2010. The Technical Specification Required Action Statement for 3.7.10 Condition B was exited on September 18, 2013.
|05000306/LER-2012-001||20 March 2013||Prairie Island||On February 21, 2012, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was performing a normal shutdown in preparation for refueling outage 2R27. With Unit 2 at approximately 11.42% power, the reactor was manually tripped in accordance with the 21/22/23 Feedwater Heater Hi-Hi alarm response procedure. There were no other unusual / not understood events associated with the shutdown. The Equipment Cause Evaluation (ECE) determined that the third stage low-pressure FWH bypass line to the condenser is potentially restricting flow to the dump valve. The ECE also determined that the Moisture Separator Reheater Control valves are ramped closed earlier than necessary and are fully closed by approximately 20% reactor power. This causes excessive moisture in the extraction steam at low power and results in more water accumulating in the low-pressure feedwater heaters. "The RCE determined that station implementation of the Corrective Action Program (CAP) and Equipment Reliability processes have not minimized risk of event recurrence while implementing long-term corrective actions by requiring interim/mitigating actions be taken prior to the next potential opportunity for recurrence.|
|05000282/LER-2011-001||7 October 2011||Prairie Island|
The causal evaluation determined that the 121 MDCLP autostart was due to operation of the in service 21 Motor Driven Cooling Water Pump at near run out pump curve condition. When 11 Containment and Auxiliary Building Chiller tripped on low evaporator temperature (from being operated under low load conditions) the Cooling Water System cooling water demand increased when the containment fan coil cooling water valves switched to the Cooling Water System. The increased cooling water flow demand dropped cooling water header pressure down to the autostart setpoint of 121 Cooling Water Pump. 121 Cooling Water Pump started as designed.
Corrective actions to resolve the issue Include performing a Cooling Water System review to determine methods and any single point vulnerabilities that can be performed to minimize the potential for
Operating procedures will also be evaluated to determine If procedural or operation period changes can be made to reduce the likelihood of autostarting a Cooling Water Pump. - NRC FO Ni 366 (10-2010) NR -20 C FORM 306A LICENSEE EVENT REPORT (LER) (110)
|05000282/LER-2011-002||23 August 2011||Prairie Island|
On July 1, 2011, Prairie Island Nuclear Generating Plant (PINGP) Unit 1 was manually tripped from approximately 100% steady-state power. The manual reactor trip was in response to the right main turbine stop valve failing closed as the result of an Electro-Hydraulic (EH) oil leak located at the stop valve. As designed, 11 and 12 Auxiliary Feed Water (AFW) pumps auto started on steam generator low water level.
The causal evaluation determined that right main turbine stop valve failed closed due to excessive oil leakage from a failed 0-ring. An age related compression set and slight extrusion near the outside contact surface resulted in the failed 0-ring.
|05000306/LER-2011-003||23 August 2011||Prairie Island|
On June 27, 2011, the off-site AC power sources to Unit 2 Prairie Island Nuclear Generating Plant (PINGP) were declared inoperable as a result of Transformer 2RY lockout and less than the required minimum voltage on the transmission system. The paths to the transmission system was declared inoperable. Although inoperable, transmission system sources remained connected to Unit 2; emergency diesel generators were available but not required to run.
The overcurrent ground detection relay actuated due to a bus phase to ground fault resulting from failed gasket material resulting in the Transformer lockout. It was determined that a less than adequate review of the Preventive Maintenance (PM) Deferral Process delayed the bus duct inspections that would have likely identified and corrected the deficient gasket material.
|05000306/LER-2011-002||30 June 2011||Prairie Island|
On May 9, 2011, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 tripped from approximately 100% steady-state power as a result of a turbine trip from a generator lockout signal. At the time of the trip Unit 1 was in a planned refueling outage in Mode 6 Operation and thunderstorms were in the vicinity of the plant.
Post event diagnostics determined that the relay ground potential circuit had a missing link (wire) between substation panels which resulted in the loss of the reference ground potential to the associated protective relaying in the Switchyard panels. This resulted in the improper operation of the breaker protection relays.
The causal evaluation determined that the proximate cause of the relaying scheme failure was inadequate ground wire installation during a modification to the substation equipment. The evaluation also determined that NSPM (Transmission and Nuclear) did not establish the appropriate standard for procedure quality with respect to work under System Control Center jurisdiction in the PINGP substation such that a means of effectively utilizing human performance tools to minimize errors during work, which directly impacts the safe and reliable operation of nuclear plants, was not in place.
|05000306/LER-2011-001||20 April 2011||Prairie Island|
While Unit 2 was operating at 100% power, improper control of Unit 2 Shield Building access doors (Doors 172 and 173) during planned maintenance resulted in an unplanned Limiting Condition for Operation (LCO) entry and a Loss of Safety Function (LOSF) for the Shield Building system. During the maintenance work, the doors were simultaneously opened several times, for approximately ten seconds each time, resulting in a LOSF. This was reportable per 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of the safety function that was needed to control the release of radioactive material.
The causal evaluation determined that Planning did not identify within the Work Order (WO) the latent impact of the LOSF when two Shield Building access doors were to be simultaneously opened.
|05000282/LER-2010-004||28 January 2011||Prairie Island|
Safety related battery chargers have the potential to stop providing an output, or "lock up", if their AC input voltage drops below their nameplate minimum voltage of 90% of 480V at the battery charger Motor Control Center. Exact voltage, duration of voltage dip, and charger loading conditions which cause lock up of chargers are unknown.0If a reduction in input voltage results in a battery charger locking up, the battery charger will not be able to recharge the battery from a partially discharged state unless operator actions are taken to restore the battery charger.
On October, 22, 2010, an operability determination concluded that specific Design Basis Accident scenarios may include this undervoltage condition which could cause the battery chargers to lock up. The design basis requirements of the safeguards battery are such that it can carry expected shutdown loads for the DC system for one hour. Compensatory measures in the form of new operator actions have been put in place to support restoration of the battery chargers within this timeframe.
|05000282/LER-2009-006||8 April 2010||Prairie Island|
On October 20, 2009, Prairie Island Nuclear Generating Plant (PINGP) Unit 1 was in Mode 5 during a refueling outage. Unit 2 was in Mode 1 operating at full power. PINGP staff determined that in the event of flooding in the turbine building, due to a postulated high energy line break (HELB), operability of the Unit 1 Emergency Diesel Generators (D1 and D2) may not be assured.
The as-found condition was an original design issue that was determined to be reportable during preparation of an analysis of flooding due to a postulated HELB in the turbine building. Physical and procedural changes have been made to minimize challenges to plant equipment and personnel in combating potential flooding events. The results and corrective actions of the ongoing evaluation of turbine building flooding due to a potential HELB will be provided in a supplemental LER.
|05000282/LER-2009-003||22 May 2009||Prairie Island|
On March 23, 2009, at approximately 1200 CDT, Northern States Power Company - a Minnesota corporation (NSPM) design engineering personnel discovered a potential vulnerability with Prairie Island Nuclear Generating Plant (PINGP) Unit 1 component cooling (CC) piping. The potential exists for a tornado missile hazard to impact the CC piping that supplies the 122 spent fuel pool (SFP) heat exchanger (HX). In the normal configuration, a missile impact to this piping would affect both trains of PINGP Unit 1 component cooling thus rendering the CC system inoperable.
This condition is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.73(a)(2)(v)(A) as a condition that could have prevented the fulfillment of a safety function of a system that is needed to maintain the reactor in a safe shutdown condition.
The condition was caused by a design change that added the 122 SFP HX prior to initial plant startup.
The 122 SFP HX is currently isolated to protect the CC system function from a tornado missile hazard.
|05000306/LER-2008-002||19 December 2008||Prairie Island|
On October 30, 2008, at 1414 CDT during reactor physics testing following a Unit 2 outage, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was just above the point of adding heat (POAH) when an urgent failure alarm was received while moving Bank A control rods inward. It was noted that Group 1 control rods in Power Cabinet 21AC unexpectedly stopped moving while Group 2 control rods in Power Cabinet 22AC continued inward rod motion. Operators stopped Group 2 control rods and initiated a manual reactor trip. All systems operated as expected following the trip and operator response and recovery actions were as expected.
Subsequent troubleshooting identified that a Phase C fuse in the 21AC Moveable Gripper bus duct disconnect switch had blown. Troubleshooting also identified that the blown fuse was not fault related and likely due to a random failure. All 21AC moveable gripper bus duct disconnect switch fuses were replaced on October 31st, 2008 at approximately 0031 CDT thereby restoring full functionality of bank A control rods.0In addition, all of the moveable gripper bus duct disconnect-switch fuses for the other two power cabinets (22AC and 21BD) in Unit 2 were replaced to ensure that the extent of condition was corrected.
Reactor startup and physics testing were resumed on October 31, 2008 at 1115 and Unit 2 reactor was returned to criticality on October 31, 2008 at 1311 CDT.