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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 54012 | 20 April 2019 10:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Both Reactor Feed Pumps Trip Causing Manual Reactor Scram and Eccs Injection | At 0507 (CDT on April 20, 2019), the DAEC (Duane Arnold Energy Center) experienced a trip of both reactor feed pumps. Operators inserted a manual scram. All control rods inserted, as required. As a result of the feed pump trips and scram, HPCI and RCIC automatically injected. Also, containment isolations occurred, as expected for this event. All systems responded as designed. Operators are currently taking the unit to cold shutdown conditions. Vessel level is being controlled by RCIC with Condensate System available. Pressure is being controlled using Main Steam Line drains and the Main Condenser is available. Normal electrical lineup remains. The cause of the reactor feed pumps tripping is believed to be an instrument air leak to flow control valves, causing loss of suction to both feed pumps. The licensee notified the NRC Resident Inspector. | Main Steam Line Main Condenser Control Rod | |
| ENS 53676 | 19 October 2018 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Trip Due to Feedwater Regulating Valve Failing Closed | At 1725 CDT, a Feedwater Regulating valve failed closed, resulting in a reactor level transient, which initiated a reactor trip, Primary Containment Isolation System signals to valves in Groups 2, 3, and 4 and initiation of High Pressure Coolant Injection and Reactor Core Isolation Cooling. All control rods inserted and level has been restored to normal. The cause of the feedwater valve failure is under investigation. All other systems responded as expected. This report is being made under 10 CFR 50.72 (b)(2)(iv)(B), (b)(3)(iv)(A) and (b)(2)(iv)(A). The Senior Resident Inspector has been informed. Decay heat is being removed via the main condenser and reactor vessel water level is being maintained by the condensate and feedwater systems. | Feedwater High Pressure Coolant Injection Primary Containment Isolation System Reactor Core Isolation Cooling Main Condenser Control Rod | |
| ENS 51484 | 20 October 2015 23:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Diesel Generators Started After Transmission Line Lightning Strikes | This report is being made under 10CFR50.72(b)(3)(iv)(A) 'any event or condition that results in a valid actuation of any system listed in paragraph (b)(3)(iv)(B) of this section' due to the automatic start of 'A' and 'B' Standby Diesel Generators. The automatic start of the Standby Diesel Generators occurred on a valid bus under-voltage condition caused by severe weather (lightning strikes) in the DAEC (Duane Arnold Energy Center) area. The Standby Diesel Generator output breakers did not close onto their respective essential buses. Essential electrical buses remained powered from off-site circuits through their normal power supply transformer, 1X3 Startup Transformer, during and after the event. The Standby Diesel Generators have been returned to the standby readiness condition. Both credited off-site power circuits were available before the event and remain available after the event. The licensee notified the NRC Resident Inspector. | 05000331/LER-2015-005 | |
| ENS 50246 | 1 July 2014 00:13:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auto-Start of Standby Diesel Generators Due to a Grid Disturbance | This report is being made under 10CFR50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section' due to an auto start of 'A' and 'B' Standby Diesel Generators. At 1913 CDT, a grid disturbance during a local thunderstorm caused a valid bus undervoltage condition that resulted in the auto start of both the 'A' and 'B' Standby Diesel Generators. The 'A' and 'B' Standby Diesel Generator supply breakers did not close onto their respective buses as they remained powered by their normal power supply, the 1X003 Startup Transformer, during and after the event. NextEra Duane Arnold personnel inspections revealed no issues with breakers in the switchyard. ITC (Iowa Transmission Company) reported a fault on a 161 Kv line in the NextEra Duane Arnold vicinity. Offsite power remained operable during and following the event. As designed the 'A' and 'B' Emergency Service Water systems auto started when the diesel generators auto started. The Emergency Service Water systems and both Standby Diesel Generators have been returned to standby/readiness condition. All ECCS (Emergency Core Cooling System) systems were available before the event and have remained available following the event. The licensee notified the NRC Resident Inspector. | Service water | 05000331/LER-2014-005 |
| ENS 47143 | 11 August 2011 16:38:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Primary Containment Isolations | At 1138 CDT, while in the process of shutting down as required by Technical Specifications (Reference EN# 47142), with the reactor at approximately 15 percent power, a manual scram was inserted in order to complete the TS Required Action of being in Mode 3 within 12 hours. Upon inserting the manual scram, reactor water level dropped below 170 inches resulting in Primary Containment Isolation System (PCIS) Groups 2, 3 and 4 being received. This reactor water level response is considered normal following a reactor scram from power due to void collapse in the reactor vessel. Reactor water level is currently being controlled in the normal band. All PCIS group isolations went to completion and were subsequently reset. The PCIS isolations all functioned properly. This event is being reported pursuant to 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. | Primary Containment Isolation System Primary containment | 05000331/LER-2011-002 |
| ENS 45873 | 26 April 2010 08:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Scram Due to High Turbine Vibrations | This report is being made under 50.72(b)(2)(iv)(B) for inserting a manual reactor scram due to rising vibrations on the #6 turbine bearing. A planned reactor shutdown was in progress with reactor power at 13.8% when turbine vibrations approached procedural limits which would require a manual scram of the reactor. The scram was uncomplicated; all control rods fully inserted. The reactor is in Mode 3, Hot Shutdown. Cooldown has been established to the condenser using main steam line drains. The NRC Resident Inspector has been informed. | Main Steam Line Control Rod | |
| ENS 45603 | 2 January 2010 05:10:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Diesel Generator Auto Start Due to Failed Line Arrestor in Switchyard | This report is being made under 10CFR50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in Paragraph (b)(3)(iv)(B) of this section' due to an auto start of the 'A' Standby Diesel Generator. The auto start of the 'A' Standby Diesel Generator occurred on a valid bus undervoltage condition caused by a momentary fault on the 161 kV Vinton line. Switchyard inspection revealed a failed 161 kV line arrestor (on) one of the phases of the Vinton line. The 'A' Standby Diesel Generator supply breaker was not required to close onto its respective essential bus as it remained powered from its normal power supply, 1X003 Start-up Transformer, during and after the event. Offsite power remained fully operable during and following the event. The 'A' Standby Diesel Generator has been returned to the standby/readiness condition. As designed, the 'A' Emergency Service Water systems auto started when the 'A' Standby Diesel Generator started. The Emergency Service Water systems have been returned to the standby/readiness condition. The 'B' Well Water Pump tripped as a result of the electrical transient. The 'A' Well Water pump and 'B' Emergency Service Water pump were manually started by operators in accordance with Abnormal Operating Procedure 408, Well Water System Abnormal Operation. The in-service Reactor Water Cleanup pump tripped which removed RWCU from service. The RWCU system has been returned to service. The in-service Spent Fuel Pool Cooling pump tripped as a result of the electrical transient. Spent Fuel Pool Cooling has been returned to service. Spent fuel pool temperature rose 0.7 (degrees) F while the pump was out of service. All ECCS systems were available before the event and have remained available following the event. The licensee is not in any Technical Specification LCO's as a result of this event. The in-service Spent Fuel Pool Cooling pump was OOS for approximately 58 minutes. The NRC Resident Inspector has been notified. | Service water Reactor Water Cleanup | 05000331/LER-2010-001 |
| ENS 45421 | 8 October 2009 19:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram During Testing | This report is being made under CFR 50.72(b)(2(iv)(B) for a automatic reactor scram which is preliminarily believed to have been caused while restoring reactor pressure vessel level/pressure instrumentation to service during performance of surveillance test 3.3.3.2-09 Reactor Water Level and Pressure Instruments Calibration . In addition, this report is being made under CRF 50.72(b)(3)(iv)(A) due to PCIS (Primary Containment Isolation System) Groups 2, 3 and 4 being received when reactor water level dropped below 170 due to the expected level shrink immediately following a reactor scram. All isolations went to completion. Reactor water level was recovered to normal vessel level as expected under post scram conditions. The following Technical Specification LCOs (Limiting Conditions of Operation) were in affect at the time of the reactor scram: TS 3.7.5, Required Action A.1, 30 day completion time for 'B' Control Building Chiller; TS 3.7.2, Required Action A.1, 7 day completion time for 'B' River Water Supply loop; TS 3.3.3.1, Function 6, Required Action A.1, 30 day completion time for position indication of MO-1949B, 'B' Residual Heat Removal Heat Exchanger Vent; TS 3.7.1, Required Action A.1, 30 day completion time for 1P22D, 'D' Residual Heat Removal Service Water Pump. All rods fully inserted during the scram. The plant is in a normal shutdown electrical lineup with offsite sources powering the safety busses. No safety or relief valves lifted during the transient. Decay heat is being removed via the steam bypass (valves to the) condenser. The licensee notified the NRC Resident Inspector. | Primary Containment Isolation System Reactor Pressure Vessel Residual Heat Removal Residual Heat Removal Service Water | |
| ENS 45154 | 23 June 2009 22:04:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auto Start of Emergency Diesel Generators Due to Offsite Voltage Transient | This report is being made under 10CFR50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(A) of this section' due to the auto start of the 'A' and 'B' Standby Diesel Generators. The auto start of the 'A' and 'B' Standby Diesel Generators occurred on a valid bus undervoltage condition caused by a transient on the 161 kV Fairfax line while there were severe thunderstorms in the DAEC area. The Standby Diesel Generator supply breakers did not close onto their respective essential buses. The essential buses were not required to load onto either diesel as they remained powered from their normal power supply, 1X003 Start-up Transformer, during and after the event. Offsite power remained fully operable during and following the event. The Standby Diesel Generators have been returned to the standby/readiness condition. As designed, both Emergency Service Water systems auto started when the SBDGs started. The Emergency Service Water systems have been returned to the standby/readiness condition. All ECCS systems were available before the event and have remained available following the event. The NRC Resident Inspector has been informed of the automatic SBDG start. | Service water Emergency Diesel Generator | |
| ENS 44965 | 3 April 2009 05:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Increasing Reactor Water Level During Calibration | This report is being made under (10 CFR) 50.72(b)(2)(iv)(B) for inserting a manual reactor scram due to indicated reactor water level of 205 inches and rising. The exact cause of the reactor water level increase is not known at this time. However, STP (Surveillance Test Procedure) 3.3.3.2-09, 'Reactor Water Level and Pressure Instruments Calibration' was in progress at the time and is likely the cause of the level transient. In addition, this report is being made under (10 CFR) 50.72(b)(3)(iv)(A) due to PCIS (Primary Containment Isolation System) groups 2, 3 and 4 being received when reactor water level dropped below 170 inches following the reactor scram. All isolations went to completion. This level response is considered normal following a reactor scram from power and reactor water level is currently being controlled in the normal band. All PCIS group isolations were reset. All rods fully inserted during the scram. The plant is in a normal shutdown electrical lineup with offsite sources powering the safety busses. No safety or relief valves lifted during the transient. Decay heat is being removed via the steam bypass to condenser valves. The licensee has notified the NRC Resident Inspector. | ||
| ENS 44821 | 2 February 2009 00:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Loss of Condenser Cooling | (The licensee manually inserted a reactor scram) due to imminent loss of heat sink (condenser). Circulating water to the condenser suffered an unisolable break resulting in loss of suction source to pumps with cavitation. (The) plant was stabilized per emergency operating procedures. Containment isolations signals operated per design with Groups 2, 3, and 4 received. Plant response (was) complicated with (an) unexpected loss of one 161 KV bus in the switchyard. This was caused by a breaker failure lockout of the bus caused by signal from the output breaker. This output breaker did open. Cause of the breaker failure being investigated. (The) power loss (of the 161 KV bus) had minor plant effect. One power supply to essential service offsite transformer 1X3 (was) lost due to the bus lockout. (The) transformer and offsite circuit remained available. (The) plant did sustain a temporary loss of spent fuel pool cooling system. Cooling has been restored with no level or temperature complications. Essential power (remained) available with both onsite SBDG's (Standby Diesel Generators) and offsite circuits (current source). All ECCS available. All safety related equipment operat(ed) as designed, except Reactor Core Isolation Cooling (RCIC) turbine which had difficulty in automatic control. Manual control (of RCIC) was effective. This is not unexpected during the conditions it was operating (low flow). The licensee was shutting down to enter a refueling outage as the time of the event. Circulating water to one cooling tower had been isolated when a riser on the second cooling tower ruptured and resulted in loss of supply to the circwater pumps. All systems functioned as required except for the lockout of the 161KV bus and RCIC automatic control. Rods fully inserted. The plant is in a normal electrical configuration except for the loss of the 161 KV bus which does not supply any safety related busses. No safety relief valves lifted during the transient. Cooling is currently via HPCI and RCIC to the torus with torus cooling via essential service water. The plant is currently stable at 600 psi and being cooled-down to cold shutdown. The NRC Resident Inspector has been notified. | Service water Reactor Core Isolation Cooling Safety Relief Valve | |
| ENS 43817 | 1 December 2007 20:35:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Edg Starts on 161 Kv Line Transient | This report is being made under 10CFR50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section' due to the Auto Start of the 'A' and 'B' Standby Diesel Generators. The Auto start of the 'A' and 'B' Standby Diesel Generators occurred on a valid bus undervoltage condition caused by a transient on the 161 kV Fairfax line. The Standby Diesel Generator Feeder Breakers did not close onto their respective Essential buses. The Essential buses were not required to load onto either diesel as they remained powered from the 1X003 Start-up Transformer during and after the event. Offsite power remained fully operable during and following the event. The Standby Diesel Generators have been returned to the standby/ready condition. The RWCU pump tripped which removed RWCU from service. The system has been returned to service. The 'D' Well Water Pump also tripped and auto restarted. The licensee notified the NRC Resident Inspector. | 05000331/LER-2007-011 | |
| ENS 43271 | 2 April 2007 16:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Lowering Reactor Vessel Water Level | This report is being made under 50.72(b)(2)(iv)(B) for inserting a manual scram due to lowering reactor pressure vessel level. The lowering level resulted from a loss of non-essential 4160 VAC bus 1A2 and subsequent loss of 'B' Reactor Feed Pump. In addition this report is being made under 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B) of this section due to PCIS groups 2, 3, and 4 being received when reactor water level dropped below 170". All isolations went to completion. The reactor water level drop is normal following a scram from full power due to void collapse in the reactor vessel. After the initial scram, vessel level rose to 211" causing the 'A' Reactor Feed Pump to trip. Before a Reactor Feed Pump could be restarted and adequate feed flow established, another scram signal and isolation occurred when vessel level dropped below 170". Control rods were already fully inserted and isolation valves in the isolated condition when this occurred. Reactor water level was restored to normal and the PCIS group isolations were reset. The current plan is to remain in Hot Standby (Mode 3) and determine the cause of the loss of 1A2 non-essential 4160 VAC bus. The licensee informed the NRC Resident Inspector.
This update is being added to clarify that the second scram on 170 " vessel level described in the original report also constituted a 50.72 (b)(3)(iv)(A) reportable event. The licensee notified the NRC Resident Inspector. Notified R3DO (Kozak) | Reactor Pressure Vessel Control Rod | 05000331/LER-2007-007 |
| ENS 43247 | 19 March 2007 00:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Coolant Chemistry Concerns | This report Is being made under 50.72(b)(2)(iv)(B) for inserting a manual reactor scram due to high levels of reactor coolant chlorides, sulfates, and conductivity. These elevated chemistry parameters may be due to a possible resin intrusion following the placement of a Condensate Demineralizer in service. In addition, this report is being made under 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B) of this section' due to PCIS groups 2, 3, and 4 being received when reactor water level dropped below 170 inches. All isolations went to completion. The reactor water level drop is normal following a scram from 28% power due to void collapse in the reactor vessel. Reactor water level was restored to normal and the PCIS group isolations were reset. The current plan is to cooldown the plant and enter Mode 4. The scram was characterized as uncomplicated. All rods fully inserted. No safety relief valves lifted. Reactor water low level was approximately 165 inches and was restored with normal reactor feedwater. Decay heat is currently being removed via steam line drains. There were no electrical alignment or grid issues as a result of the transient. No significant Technical Specification LCOs were in effect at the time of the scram. The licensee notified the NRC Resident Inspector. | Feedwater Safety Relief Valve | 05000331/LER-2007-006 |
| ENS 43210 | 3 March 2007 05:32:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Signal Due to High Discharge Volume Level | The licensee was performing testing that inserts a reactor scram manually to verify that the backup scram valves port air. Moments after the scram was reset a second scram came in from a scram discharge volume high level. The second scram was not called out in the test, nor were there steps to bypass the scram discharge volume scram signal prior to resetting the manual scram. All rods were already fully inserted and, consequently, there was no rod motion as a result of the scram. The licensee states the unexpected scram signal was a result of an error in the test procedure. The NRC Resident Inspector has been notified. | 05000331/LER-2007-005 | |
| ENS 43185 | 24 February 2007 23:56:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Normal Electrical Lineup Due to Grid Instability from Adverse Weather Conditions | Immediate notifications are being made for the following events: 1) Valid RPS non-critical actuation: 8 hour notification per 50.72(b)(3)(iv)(B) 2) Group 1, 2, 3, 4 and 5 isolations: 8 hour notification per 50.72(b)(3)(iv)(B) 3) Loss of decay heat removal (RHR Shutdown Cooling) isolated: 8 hour notification per 50.72(b)(3)(v)(A) 4) Auto start of 'A' Standby Diesel Generator and subsequent pick up of essential buses by both 'A' and 'B' Standby Diesel Generators: 8 hour notification per 50.72(b)(3)(iv)(A). Just prior to the event the plant was in Mode 5 with core alterations in progress and the 'B' Standby Diesel Generator being run manually for post maintenance testing. At 1756 hours on 2/24/2007, core alterations were halted due to indications that the grid was becoming unstable. Freezing rain had been in progress all day. At 1757 hours an electrical transient occurred as some of the offsite lines coming into the switchyard were lost. The 'B' Reactor Protection System (RPS) bus was lost. A full RPS trip occurred from loss of the 'B' RPS bus and too few inputs to the 'A' RPS trip logic. The loss of 'B' RPS power resulted in Groups 1, 2, 3, 4 and 5 isolations. The Group 4 isolation caused a loss of decay heat removal capability (RHR Shutdown Cooling isolated). The 'A' Standby Diesel Generator auto started, but the 1A3 essential bus remained powered from offsite sources. The 'B' SBDG was already running. The 1A4 bus also remained powered from offsite sources. At 1820 hours, a degraded voltage trip signal was received resulting in both emergency diesel generators picking up their respective buses. At present, the essential buses 1A3 and 1A4 are being carried by the emergency diesel generators. Only one of the six lines supplying offsite power to the site is available. Both non-essential buses are being powered from the one offsite line. The emergency buses will remain on the emergency diesel generators until additional offsite lines can be restored. RHR Shutdown Cooling was restored at 1826 hours. The licensee notified the NRC Resident Inspector, State and local officials. | Reactor Protection System Emergency Diesel Generator Shutdown Cooling Decay Heat Removal | |
| ENS 42966 | 6 November 2006 07:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram During Turbine Testing | On 6 November 2006 at 0110 an automatic reactor scram was received. This report is being made 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.' The plant also received Group 2, 3 and 4 isolations from the low level signal as reactor water dropped due to the reactor scram. The isolations are being reported under 10 CFR 50.72 (b)(3)(iv)(A), 'any event or condition that results actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section.' The scram is believed to be the result of a turbine trip but is currently being investigated further. Main Turbine surveillance testing was in progress at the time but it is not known if the testing caused the turbine trip and subsequent Reactor Scram. Reactor pressure is currently being controlled by the turbine bypass valves and steamline drains. No actuations of Safety Relief Valves were required or occurred. Both Reactor Recirculation pumps tripped as part of RPT. The 'B' Recirc pump has been restarted. Reactor level dropped following the scram resulting in the isolations but was recovered and is currently being maintained by the normal feedwater systems. All isolations went to completion and have been reset at this time. The plant electrical buses transferred without incident and are currently in a normal shutdown alignment. The current plan is to remain in Mode 3 while investigations determine the cause of the turbine trip and reactor scram. All control rods fully inserted on the trip. Currently the plant is using offsite power but diesel generators are available. The licensee notified the NRC Resident Inspector. | Feedwater Reactor Protection System Main Turbine Reactor Recirculation Pump Safety Relief Valve Control Rod | |
| ENS 40353 | 25 November 2003 21:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram Due to Degrading Condenser Vacuum During Power Ascension | Unit was in process of returning to operation following repair of condenser expansion joint. During power ascension, a manual reactor scram was inserted due to degrading condenser vacuum. During the reactor water level transient after the scram, reactor water level dropped below 170 inches and caused the isolation of PCIS groups 2, 3, and 4. All isolations went to completion. This reactor water level drop is normal following a scram from 20% power due to void collapse in the reactor vessel. Reactor water level was restored to normal using the feedwater system and the PCIS group isolations are being reset. Current plan is to remain in Mode 3 during condenser vacuum troubleshooting and repair. All rods were fully inserted during the scram, no SRVs lifted, and the electric distribution system is in a normal lineup for plant condition. The MSIVs are open with decay heat being removed to the main condenser using the bypass valves. No additional specified system actuations occurred. Licensee notified the NRC Resident Inspector. | Feedwater Main Condenser | 05000331/LER-2003-006 |
| ENS 40301 | 7 November 2003 07:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram at Duane Arnold Due to Hi Coolant Conductivity | Inserted a Manual Reactor Scram due to high reactor coolant conductivity. Source of high reactor coolant conductivity is not certain at this time. In addition, this report is being made under 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B) of this section' due to PCIS groups 2, 3, and 4 being received when reactor water level dropped below 170". All isolations went to completion. The reactor water level drop is normal following a scram from 45% power due to void collapse in the reactor vessel. Reactor water level was restored to normal and the PCIS group isolations were reset. The current plan is to cool down the plant and enter Mode 4, plant cool down is being achieved using the RCIC system until the source of the conductivity transient is identified and isolated. The licensee indicated shortly prior to the scram a coolant chemistry transient was in progress. The limit for reactor coolant conductivity is 5 micro mhos per centimeter and the highest conductivity reading reached was 8 micro mhos per centimeter. The licensee speculates the most likely source of the high conductivity is related to an off service condensate demineralizer. There were no automatic ECCS initiations. No SRVs lifted. All control rods properly inserted. The licensee notified the NRC Resident Inspector. | Control Rod |