Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 40625 | 30 March 2004 13:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram During Turbine Thrust Bearing Wear Detector Testing | At 0740 hours (CST) during testing of the turbine thrust bearing wear detector, a main turbine trip occurred. This resulted in an automatic reactor scram due to turbine stop valve closure. Following the scram all Group II (Primary Containment) and Group III (Reactor Water Cleanup) isolations occurred as expected. All essential equipment functioned as required. Unit 2 remains in Mode 3 with reactor water level in the normal level band. An investigation into the Unit 2 turbine trip is in progress. Unit 1 was unaffected by the event and remains at 85% power. All control rods fully inserted. Decay heat is being removed via steam to the main condenser using the bypass valves. The Licensee notified the NRC Resident inspector. | |
| ENS 40846 | 28 June 2004 06:47:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Earthquake Felt at the Site | On June 28, 2004 at 0147 CDT, the Control Room was notified that an earthquake was felt on site by Security Personnel. Both units 1 & 2 remain on line and stable with no abnormalities and no physical or structural damage identified. The Plant seismograph DID NOT trigger. An unusual event was declared at 0151 CDT. The licensee notified the NRC Resident Inspector, State and local authorities. NOTE: See related EN#40844 and EN#40845.
On June 28, 2004 at 0318 CDT, the Unusual Event has been terminated. A complete tour of all plant vital and protected areas was performed with no physical or structural damage verified. Security performed an inspection of outside areas of plant property and noted nothing unusual. The licensee notified the NRC Resident Inspector, State and local authorities. Notified by e-mail: Susan Frant (IRO), Thomas Kozak (R3 DO), Frank Gillespie (NRR EO), Pat Hiland (Reg3), and Geoffrey Grant (R3). Also notified FEMA (Erwin Casto) and DHS (Andy Akers). | |
| ENS 40912 | 31 July 2004 02:25:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Lpci Loop Select Logic Found Degraded Due to Pressure Instruments Out-Of-Tolerance | On July 30, 2004 at approximately 2125 hours CDT, Instrument Maintenance personnel were performing the Unit One Reactor Low Pressure (RHR/LPCI) Calibration and Functional Test. At this time, the technicians determined that two out of the four "Reactor Steam Dome Pressure - Low" switches in LPCI loop select logic were out of tolerance high. These switches delay completion of the LPCI loop select logic during Single Loop Reactor Recirculation Pump operations (SLO) until the reactor pressure is less than 900 psig. With these switches out of tolerance high, LPCI loop select logic sensitivity was degraded, increasing the potential to select the wrong loop for injection during periods of SLO. At the time of discovery, both Reactor Recirculation pumps were in operation. Both switches were recalibrated immediately upon discovery to within Technical Specification tolerances. The NRC Resident Inspector was notified of this event by the licensee. | |
| ENS 41106 | 9 October 2004 02:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Contol Room Emergency Ventilation System Declared Inoperable | At 2120 (hrs. CDT) on October 8, 2004, while performing the 'Control Room Emergency Ventilation System Test,' which verifies the integrity of the control room envelope, it was determined that the positive pressure requirement of greater than or equal to 0.125 inches water gauge for the control room envelope in Technical Specification Surveillance Requirement 3.7.4.4 could not be met for all specified test points. As a result, Control Room Emergency Ventilation System was declared inoperable and Technical Specification 3.7.4, Condition A was entered. Recently completed surveillance testing has demonstrated that a positive pressure ranging from 0.056 to 0.301 inches water gauge is being maintained in the control room envelope; therefore it is expected that the safety function is being met. However, this notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) because the Control Room Emergency Ventilation System is a single train safety system and the Technical Specification requirement is not met. The affect of the failure to meet the Technical Specification requirements on the ability to perform the safety function is continuing to be evaluated. The licensee notified the NRC resident inspector. | |
| ENS 41139 | 19 October 2004 23:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection (Hpci) Inoperable Due to Controller Signal Converter Trouble. | On October 19, 2004 at approximately 1830 hours CDT, Unit One was performing QCOS 2300-05, Quarterly HPCI Pump Operability Test. This was being performed to prove operability following maintenance work on various valves and the turning motor gear unit. At this time when the HPCI turbine was rolled, the HPCI Signal Converter Trouble alarm was received. The HPCI Flow Controller demand controlled at approximately 7250 gpm instead of controlling at the desired 5600 gpm. HPCI was determined not to be operable and was shutdown per procedure. Due to the unexpected behavior of the HPCI Flow Controller, at this time it is not certain if the HPCI will meet its safety function. Therefore, we are reporting this event under 10 CFR 50.72 (b)(3)(v). Licensee entered Tech Spec 3.5.1.(f) 14 day Limiting Condition of Operation (LCO) for HPCI. Reactor Core Isolation Cooling (RICI) is operable. All other Emergency Core Cooling Systems (ECCS) and the Emergency Diesel Generators (EDG) are fully operable if needed. The NRC Resident Inspector was notified of this event by the licensee.
The purpose of this report is to retract the ENS report made on October 19, 2004 at 2237 CDT (ENS #41139). The initial report was made following HPCI operability testing in accordance with QCOS 2300-05, Quarterly HPCI Pump Operability Test. When the HPCI turbine was rolled, the HPCI Signal Converter Trouble alarm was received. The HPCI Flow Controller failed to control at the desired 5600 gpm (instead, the system ramped to 7250 gpm). Due to the unexpected behavior, it was not certain at the time if HPCI could have met its design basis requirements. However, a subsequent review of this event has determined that HPCI would have performed its safety function. The Signal Converter failed in a manner that prevented automatic flow control, but did not prevent HPCI from initiating and ramping to full flow (i.e., the turbine high speed stop). In this condition, HPCI would have met corresponding Technical Specifications and Accident Analysis requirements. The circuit board was replaced and the Signal Converter and flow controller feedback loop were re-calibrated. The Unit 1 HPCI turbine's automatic flow control was successfully tested on October 20, 2004. The Licensee notified the NRC Resident Inspector. Notified R3DO (R. Garner) | |
| ENS 41225 | 25 November 2004 21:45:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Safety Parameter Display System | On November 25, as of 1545 hours, the Station's Plant Process Computer (PPC) had been in a failed state for greater than eight hours. Previous to this, at 0745 hours, the PPC experienced an unexpected hardware failure. This computer feeds the Safety Parameter Display System (SPDS) for both Units. This failure is considered a major loss of emergency assessment capability. Troubleshooting and repairs are in progress. This failure affects the transmission of ERDS data and applicable emergency response personnel have been notified of the change in their reporting status if an emergency event is declared due to the loss of ERDS. The licensee notified the NRC Resident Inspector.
At 1040 CST on 11/26/04, Information Technology personnel restored the Station's PPC to the point where the SPDS and ERDS are both fully functional for both units. The licensee notified the NRC Resident Inspector. | |
| ENS 41369 | 4 February 2005 00:19:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | 4160 Volt Relaying and Metering Single Failure Vulnerability | The licensee provided the following report via facsimile: On February 3, 2005 at 1819 hours, Quad Cities Nuclear Power Station (QCNPS) confirmed a vulnerability with a 4160 VAC relaying and metering current transformer (CT) associated with the Unit and Reserve Auxiliary Transformers (i.e., the UAT and RAT) on both Unit 1 and Unit 2. Although the CT is currently fully operable, failure of the CT circuitry could cause the neutral overcurrent relay to trip and lockout the circuit breakers supplying feeds to safety related buses 13 (23) and 14 (24), isolating them from their normal and emergency power sources. Emergency power (i.e., the emergency diesel generator) would still be available to supply power to safety related buses 13-1 (23-1) and 14-1 (24-1), but the Residual Heat Removal Service Water (RHRSW) system may be without a source of power. If this failure occurred during a LOCA, then the RHRSW pumps may not be started within the ten minute requirement. A modification is in progress to eliminate this vulnerability. This event is being reported as a potential loss of safety function (10CFR50.72(b)(3)(v)(B)) 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), and #41370 (Dresden). | |
| ENS 41524 | 24 March 2005 15:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Minimum Switchyard Voltage Requirements Not Met | The following information was obtained from the licensee via facsimile (licensee text in quotes): Minimum Switchyard Voltage Requirements Not Met On March 24, 2005, at 0930 hours, Quad Cities was notified that the switchyard voltage was below that required to ensure that offsite power would remain available following a design basis accident. Both sources of off-site power were declared inoperable. The appropriate Technical Specification required actions were taken for both units. The ability of the Emergency Diesel Generators to perform their design function is not affected by this condition. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function given the predicted post-LOCA switchyard voltage. A preliminary engineering assessment indicates that one source of offsite power was available. Additional confirmatory reviews are being performed to determine if the safety function was preserved. At this time, grid/switchyard voltage has been restored. Minimum switchyard voltage required is 348.4 KV. The analyzed minimum voltage which prompted notification to the NRC was 347.5 KV. Switchyard voltage at the time of this report was ~359 KV. The licensee has notified the NRC Resident Inspector.
The purpose of this report is to retract the following ENS reports: 1) ENS # 41524 on March 24, 2005; 2) ENS # 41562 on April 4, 2005; and 3) ENS # 41587 on April 11, 2005. These reports were made following notification to Quad Cities that the switchyard voltage was below the required value necessary to ensure that offsite power would remain available following a design basis accident. For each event, both sources of offsite power were declared inoperable, the appropriate Technical Specification required actions were taken for both units, and an ENS notification was made for a condition that could have prevented fulfillment of a safety function. The events occurred when Unit 1 was in a refueling outage and Unit 2 was operating at 85% power. Subsequent Engineering analysis has determined that with a unit shutdown, and the large 4 Kv loads on that unit not running, the required minimum post-accident switchyard voltage is reduced to 339.3 kV (for the shutdown unit). Given the new minimum post-accident switchyard voltage, it was confirmed that there was a source of off-site power available to the Station. Specifically for all three events, the projected post-accident switchyard voltage (which ranged between 343 kV to 347.5 kV) was higher than the required minimum post-accident switchyard voltage (339.3 kV) for the shutdown unit. As a result, there was no condition present that could have prevented fulfillment of a safety function, and thus these events are not reportable. The licensee notified the NRC Resident Inspector. Notified R3 DO (J. Madera) | |
| ENS 41531 | 28 March 2005 01:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Electrical Safety Bus Trip During Quad Cities Unit 1 Refueling Outage | The following information was obtained from the licensee via facsimile (licensee text in quotes): At 1930, Unit 1 experienced a loss of 480 VAC busses 18 and 19. This caused a loss of power to the Control Room Emergency Ventilation System (CREVS) and the CREVS Air Conditioning System (CREVS AC). This event also caused a loss of power to the U1 equipment that was supporting the Alternate Decay Heat Removal (ADHR) mode of operation. Power was restored to Bus 19 at 2001 and Bus 18 at 2013. The restoration of Bus 18 also restored power to CREVS and CREVS AC. All systems supporting ADHR were restored by 2015. At the time of the occurrence, the estimated time to boil without decay heat removal capability was 571 minutes. All isolations and actuations occurred as expected. The cause of the bus trips is being investigated. This Event is being reported under 50.72 (b)(3)(v)(B) and 50.72 (b)(3)(v)(D). The licensee stated at the time of the event, Busses 18 and 19 were cross-tied and the feeder breaker to Bus 18 tripped open. The breaker was changed out and power to Busses 18 and 19 restored. The cause of the Bus 18 feeder breaker trip has not yet been determined. The licensee has notified the NRC Resident Inspector.
The licensee stated that the Alternate Decay Heat Removal pumps from Unit 2 remained in service so that all decay heat removal was not lost. The Unit 1 primary coolant system temperature increase during the 45 minute duration of this event was approximately 1 degree. R3DO (Kozak) has been notified. | |
| ENS 41562 | 4 April 2005 22:07:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Minimum Switchyard Voltage Requirements Not Met | The following information was obtained from the licensee via facsimile (licensee text in quotes): On April 4, 2005, at 1607 hours (CDT), Quad Cities was notified that the switchyard voltage was below that required to ensure that offsite power would remain available following a design basis accident. Both sources of off site power were declared inoperable. The appropriate Technical Specification required actions were taken for both units. The ability of the Emergency Diesel Generators to perform their design function is not affected by this condition. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function given the predicted post-LOCA switchyard voltage. This notification report is similar to the condition reported on March 24, 2005 (reference EN#41524 for additional information). Technical Specification minimum switchyard voltage is 348.4 KV. Switchyard voltage at 1607 hrs. was approximately 343.7 KV. Switchyard voltage at the time of NRC notification was 358 KV which is above the Technical Specification minimum. The licensee has notified the NRC Resident Inspector.
The purpose of this report is to retract the following ENS reports: 1) ENS # 41524 on March 24, 2005; 2) ENS # 41562 on April 4, 2005; and 3) ENS # 41587 on April 11, 2005. These reports were made following notification to Quad Cities that the switchyard voltage was below the required value necessary to ensure that offsite power would remain available following a design basis accident. For each event, both sources of offsite power were declared inoperable, the appropriate Technical Specification required actions were taken for both units, and an ENS notification was made for a condition that could have prevented fulfillment of a safety function. The events occurred when Unit 1 was in a refueling outage and Unit 2 was operating at 85% power. Subsequent Engineering analysis has determined that with a unit shutdown, and the large 4 Kv loads on that unit not running, the required minimum post-accident switchyard voltage is reduced to 339.3 kV (for the shutdown unit). Given the new minimum post-accident switchyard voltage, it was confirmed that there was a source of off-site power available to the Station. Specifically for all three events, the projected post-accident switchyard voltage (which ranged between 343 kV to 347.5 kV) was higher than the required minimum post-accident switchyard voltage (339.3 kV) for the shutdown unit. As a result, there was no condition present that could have prevented fulfillment of a safety function, and thus these events are not reportable. The licensee notified the NRC Resident Inspector. Notified R3 DO (J. Madera) | |
| ENS 41587 | 11 April 2005 18:40:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Minimum Switchyard Voltage Requirements Not Met | At 1240 on April 11, 2005 Quad Cities Station was notified calculated post-LOCA switchyard voltage is below the minimum acceptable value required to ensure offsite power will remain available following a design basis accident. The appropriate Technical Specification Actions have been taken for both Units. The ability of the Emergency Diesel Generators to fulfill their design function is not affected by this condition. This event is being reported in accordance with 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function given the predicted post-LOCA switchyard voltage. Actual Switchyard voltage at the time of this notification is 361 Kv. The required minimum post-accident switchyard voltage is 348.4 Kv, while the projected post-accident voltage which prompted this notification is 346.3 Kv. This notification is similar to the condition reported on March 24, 2005 (EN #41524) and April 4, 2005 (EN #41562). The licensee has notified the NRC Resident Inspector.
The purpose of this report is to retract the following ENS reports: 1) ENS # 41524 on March 24, 2005; 2) ENS # 41562 on April 4, 2005; and 3) ENS # 41587 on April 11, 2005. These reports were made following notification to Quad Cities that the switchyard voltage was below the required value necessary to ensure that offsite power would remain available following a design basis accident. For each event, both sources of offsite power were declared inoperable, the appropriate Technical Specification required actions were taken for both units, and an ENS notification was made for a condition that could have prevented fulfillment of a safety function. The events occurred when Unit 1 was in a refueling outage and Unit 2 was operating at 85% power. Subsequent Engineering analysis has determined that with a unit shutdown, and the large 4 Kv loads on that unit not running, the required minimum post-accident switchyard voltage is reduced to 339.3 kV (for the shutdown unit). Given the new minimum post-accident switchyard voltage, it was confirmed that there was a source of off-site power available to the Station. Specifically for all three events, the projected post-accident switchyard voltage (which ranged between 343 kV to 347.5 kV) was higher than the required minimum post-accident switchyard voltage (339.3 kV) for the shutdown unit. As a result, there was no condition present that could have prevented fulfillment of a safety function, and thus these events are not reportable. The licensee notified the NRC Resident Inspector. Notified R3 DO (J. Madera) | |
| ENS 41652 | 29 April 2005 12:32:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Minimum Switchyard Voltage Requirements Not Met | At 0732 (CDT) on April 29, 2005 Quad Cities Station was notified that the calculated post-LOCA switchyard voltage was below the minimum acceptable value required to ensure offsite power would remain available following a design basis accident. As a result, both offsite power sources were declared inoperable and the appropriate Technical Specification Actions were taken for both Units. The ability of the Emergency Diesel Generators to fulfill their design function was not affected by this condition. This event is being reported in accordance with 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function given the predicted post-LOCA switchyard voltage. The condition was exited at 0836 (CDT) on April 29, 2005. The required minimum post-accident switchyard voltage is 352.9 KV (Unit 1) and 351.0 KV (Unit 2) while the projected post accident voltage, which prompted this notification, was 347.9 KV. During this timeframe, actual switchyard voltage was approximately 358 KV. The licensee notified the NRC Resident Inspector.
The purpose of this report is to retract ENS report #41652 (April 29, 2005), ENS report 41766 (June 11, 2005), and ENS report #41801 (June 27, 2005). The reports were made following notification that the switchyard predicted voltage was below the required value necessary to ensure that offsite power would remain available following a design basis accident. For each event, both sources of offsite power were declared inoperable, the appropriate Technical Specification required actions were taken for the units, and an ENS notification was made for a condition that could have prevented the fulfillment of a safety function. Each Quad Cities unit has two sources of off-site power. For a specific unit, one source is from the switchyard through the unit's Reserve Auxiliary Transformer (RAT) and the second source is from the switchyard through the opposite unit's RAT. Following further analysis of the loading on the RATs during an accident, it was determined that only one source of offsite power to each unit should have been considered inoperable. The analysis determined that because the RAT for the non-accident unit (i.e., the offsite source supplied through the opposite unit) would be carrying a much smaller load, it would not be subject to a degraded voltage condition and resultant isolation from the grid. Therefore, an accident on one unit would not affect the operability of the source of offsite power supplied through the opposite unit's RAT. As a result, only the source of offsite power supplied through the unit's RAT was inoperable. With one source of offsite power available at all times, there was no condition present that could have prevented fulfillment of a safety function, and therefore, these events are not reportable. The licensee notified the NRC Resident Inspector. Notified R3 DO (R. Lanksbury) | |
| ENS 41766 | 11 June 2005 18:10:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Minimum Switchyard Voltage Requirements Not Met | At 1310 on June 11, 2005 Quad Cities Station was notified that the calculated post-LOCA switchyard voltage is below the minimum acceptable value required to ensure offsite power will remain available following a design basis accident. The appropriate Technical Specification Actions have been taken for both Units. The ability of the Emergency Diesel Generators to fulfill their design function is not affected by this condition. This event is being reported in accordance with 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function given the predicted post-LOCA switchyard voltage. Actual Switchyard voltage at the time of this notification is 356 Kv. The required minimum post-accident switchyard voltage for current conditions is 352.9 Kv for U-1 and 351 Kv for U-2, while the projected post-accident voltage which prompted this notification is 347.4 Kv. This notification is similar to the condition reported on March 24, 2005 (EN #41524), April 4, 2005 (EN #41562) and April 11, 2005 (EN #41587). Both Units entered T.S. 3.8.1 which requires verification of EDG operability (completed) and continued monitoring of line voltages with restoration of at least one offsite-line to the required minimum within 24-hours. The licensee informed the NRC Resident Inspector.
At 1505 CDT both Units exited T.S. 3.8.1 when projected post-accident line voltage reached 353.8 kV. Notified R3DO (Lara).
The purpose of this report is to retract ENS report #41652 (April 29, 2005), ENS report 41766 (June 11, 2005), and ENS report #41801 (June 27, 2005). The reports were made following notification that the switchyard predicted voltage was below the required value necessary to ensure that offsite power would remain available following a design basis accident. For each event, both sources of offsite power were declared inoperable, the appropriate Technical Specification required actions were taken for the units, and an ENS notification was made for a condition that could have prevented the fulfillment of a safety function. Each Quad Cities unit has two sources of off-site power. For a specific unit, one source is from the switchyard through the unit's Reserve Auxiliary Transformer (RAT) and the second source is from the switchyard through the opposite unit's RAT. Following further analysis of the loading on the RATs during an accident, it was determined that only one source of offsite power to each unit should have been considered inoperable. The analysis determined that because the RAT for the non-accident unit (i.e., the offsite source supplied through the opposite unit) would be carrying a much smaller load, it would not be subject to a degraded voltage condition and resultant isolation from the grid. Therefore, an accident on one unit would not affect the operability of the source of offsite power supplied through the opposite unit's RAT. As a result, only the source of offsite power supplied through the unit's RAT was inoperable. With one source of offsite power available at all times, there was no condition present that could have prevented fulfillment of a safety function, and therefore, these events are not reportable. The licensee notified the NRC Resident Inspector. Notified R3 DO (R. Lanksbury) | |
| ENS 41770 | 15 April 2005 20:15:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation During Excess Flow Check Valve Testing | At 1515 hours on April 15, 2005, while conducting excess flow check valve testing in conjunction with a reactor vessel pressure test during a scheduled refueling outage, Unit 1 received an invalid Anticipated Transient Without Scram (ATWS) Channel A trip signal and invalid reactor Lo-Lo water level trips on the A and C channels for ECCS. The operating Reactor Recirculation pump (1 A) tripped and both Emergency Diesel Generators (Unit 1 and Unit 0) auto-started and ran unloaded. RCIC received an initiation signal, which resulted in proper valve repositioning for the system but no injection due to the steam supply being isolated to support the reactor vessel pressure test. In addition, all appropriate running equipment tripped due to load shedding. Reactor pressure was 787 psig at the time of the event and no pressure transient or vessel injection occurred. All equipment responded to the actuation as expected for plant conditions. The excess flow check valve testing was stopped and the reactor vessel pressure test was terminated as a result of this invalid actuation. It was subsequently determined that a leaking drain valve at the instrument rack where the excess flow check valve testing was being performed caused a slow depressurization of the high-pressure side sensing line for the level transmitters associated with ATWS Channel A and the A and C channels for ECCS, resulting in an erroneous level signal and the invalid actuation. The instrument rack drain valve is scheduled for replacement. The NRC Resident Inspector was notified | |
| ENS 41782 | 17 June 2005 16:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Ehc Malfunction | At 1120 hours (CDT) on June 17, 2005, the Unit 1 reactor automatically scrammed from 85% power due to a valid high reactor pressure signal. The maximum reactor pressure was approximately 1044 psig during the event. All control rods inserted to their full-in position. Initial indications are that the reactor pressure increase was caused by a malfunction in the Electro-Hydraulic Control (EHC) system, which resulted in closure of the main turbine control valves. The main turbine bypass valves (nine) opened as expected in response to the pressure increase. No reactor pressure vessel safety or relief valves were required to actuate during the event. Reactor water level decreased to approximately -20 inches, which resulted in automatic Group 2 and 3 isolations as expected. All systems responded properly to the event. Unit 1 is in Mode 3 with a cooldown in progress and reactor water level in the normal level band. An investigation into the Unit 1 scram is in progress. Unit 2 remains at 94% power. This report is being made in accordance with 10CFR 50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector. | |
| ENS 41801 | 27 June 2005 16:50:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Minimum Switchyard Voltage Requirements Not Met | At 1150 on June 27, 2005 Quad Cities Station was notified that the calculated post-LOCA switchyard voltage is below the minimum acceptable value required to ensure offsite power will remain available following a design basis accident. The appropriate Technical Specification Actions have been taken for both Units. The ability of the Emergency Diesel Generators to fulfill their design function is not affected by this condition. This event is being reported in accordance with 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function given the predicted post-LOCA switchyard voltage. Actual Switchyard voltage at the time of this notification is 356 Kv. The required minimum post-accident switchyard voltage for current conditions is 352.9 Kv for Unit-1 and 351 Kv for Unit-2, while the projected post-accident voltage which prompted this notification is 349.8 Kv. At 1320 hrs on June 27, the projected post-accident voltage was 351.3kV. This prediction satisfies the minimum Unit 2 voltage requirement, but is still below the minimum voltage for Unit 1. This notification is similar to the condition reported on April 29, 2005 (EN #41652) and June 11, 2005 (EN #41766). The licensee notified the NRC Resident Inspector.
The purpose of this report is to retract ENS report #41652 (April 29, 2005), ENS report 41766 (June 11, 2005), and ENS report #41801 (June 27, 2005). The reports were made following notification that the switchyard predicted voltage was below the required value necessary to ensure that offsite power would remain available following a design basis accident. For each event, both sources of offsite power were declared inoperable, the appropriate Technical Specification required actions were taken for the units, and an ENS notification was made for a condition that could have prevented the fulfillment of a safety function. Each Quad Cities unit has two sources of off-site power. For a specific unit, one source is from the switchyard through the unit's Reserve Auxiliary Transformer (RAT) and the second source is from the switchyard through the opposite unit's RAT. Following further analysis of the loading on the RATs during an accident, it was determined that only one source of offsite power to each unit should have been considered inoperable. The analysis determined that because the RAT for the non-accident unit (i.e., the offsite source supplied through the opposite unit) would be carrying a much smaller load, it would not be subject to a degraded voltage condition and resultant isolation from the grid. Therefore, an accident on one unit would not affect the operability of the source of offsite power supplied through the opposite unit's RAT. As a result, only the source of offsite power supplied through the unit's RAT was inoperable. With one source of offsite power available at all times, there was no condition present that could have prevented fulfillment of a safety function, and therefore, these events are not reportable. The licensee notified the NRC Resident Inspector. Notified R3 DO (R. Lanksbury) | |
| ENS 42179 | 30 November 2005 21:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Ventilation Ac System Inoperable | On November 30, 2005, at 1500 hours, the Control Room Emergency Ventilation Air Conditioning (CREV AC) system was declared inoperable due to failure of the associated air conditioning compressor. As a result, Technical Specification 3.7.5 Condition A was entered. The CREV AC system maintains a habitable control room environment and ensures the operability of components in the control room emergency zone during accident conditions. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) because the CREV system is a single train system, and loss of the CREV AC could impact the plant's ability to mitigate the consequences of an accident. The licensee informed the NRC Resident Inspector. | |
| ENS 42260 | 15 January 2006 17:04:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Ts Shutdown Due to Electromatic Relief Valves (Ervs) Declared Inoperable | At 1014 hours on January 15, 2006, the four Unit 1 Electromatic Relief Valves (ERVs) were declared inoperable due to concerns related to the ERV actuators. These operability concerns originated from inspections and testing performed on the Unit 2 ERVs which indicate evidence of mechanical interference issues that could prevent proper ERV operation (note that Unit 2 was shutdown on January 14, 2006 to perform ERV actuator inspections). In accordance with Technical Specification 3.4.3, Condition B, 3.5.1 Condition H, and 3.6.1.6 Condition B, Unit 1 commenced a shutdown on January 15, 2006 at 1104 hours. This notification is being made in accordance with 10 CFR 50.72(b)(2)(i) as commencement of any nuclear plant shutdown required by the plant's Technical Specifications. During the upcoming Unit 1 outage, actuator inspections and ERV testing is planned to confirm the operational ability of the ERVs. Note that the Unit 1 Target Rock Relief Valve and the Main Steam Safety Valves are operable and available to support reactor overpressure protection. The plant is conducting a normal shutdown per Tech Spec requirements. At 16:00 CST the unit will going off line and by 24:00 CST the plant is expected to be in Mode 4 (Cold Shutdown). The licensee notified the NRC Resident Inspector. | |
| ENS 42356 | 22 February 2006 07:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Valid Main Turbine Load Reject Signal | At 0122 hours (CST) on February 22, 2006, the Unit 1 reactor automatically scrammed from 85% power due to a valid turbine-generator load reject signal. The maximum reactor pressure was approximately 1050 PSIG during the event. All control rods inserted to their full-in position. Reactor water level decreased to approximately -10 inches, which resulted in automatic Group 2 and 3 isolations as expected. All systems responded properly to the event. Unit 1 is in Mode 3, maintaining reactor pressure, and reactor water level in the normal level band. An investigation into the Unit 1 scram and load reject is in progress. Unit 2 remains at 85% power. This report is being made in accordance with 10CFR 50.72(b)(2)(iv)(B) and 10 CFR50.72(b)(3)(iv)(A). No SRVs opened during the transient. Post shutdown electrical lineup is normal with the exception of the loss of one offsite power supply. Decay heat is being removed via the turbine bypass valves to the main condenser and feedwater being provided via main feedwater. Unit 2 is in a Tech Spec LCO for Loss of One Offsite power source. The licensee notified the NRC Resident Inspector. | |
| ENS 42455 | 31 March 2006 03:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Feedwater Check Valve Leakage Above Allowable Limits | At 21:00 hours on 3/30/06, Unit Two feed water header check valves 2-0220-58A and 2-0220-62A LLRT (local leak rate test) results were both indeterminate, as a result both were greater than the allowable La containment leakage rate allowed by Tech Spec 5.5.12." La is defined as the maximum allowable leak rate at a specified pressure. This is reportable under 10CFR50.72(b)(3)(ii). The "indeterminate" leak rate was higher than the test equipment could read. The licensee notified the NRC Resident Inspector.
The licensee is retracting this report based on the following: ENS Retraction - Feed Water Header Check Valve Leakage > La Containment Leakage The purpose of this notification is to retract the ENS report made on 3/31/2006 at 0226 EDT (Event # 42455). The initial report was made when it appeared the Unit 2 feed water header check valves 2-0220-58A and 2-0220-62A LLRT (local leak rate test) results were both indeterminate. As a result, both valves appeared to have had a leakage rate greater than the allowable La containment leakage rate allowed by Tech Spec 5.5.12, where La is defined as the maximum allowable primary containment leak rate at accident pressure. Quad Cities assigned all leakage experienced during this test to the primary containment isolation valves being tested and reported the situation as a condition that seriously degraded the nuclear power plant per 10CFR50.72(b)(3)(ii)(A). Subsequently, Quad Cities Station performed testing on 4/2/06 utilizing a higher fill rate test methodology. The subsequent test was completed with satisfactory results. Based on this subsequent testing, ENS report (# 42455) is being retracted. The licensee informed the NRC Resident Inspector. Notified R3DO (Pelke). | |
| ENS 42574 | 14 May 2006 14:57:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Automatic Start | During restoration from a scheduled LCO, the Unit 1 EDG auto started when its control switch was placed in the auto position. The auto start occurred as a result of a 4KV Bus Feed Breaker being open. This breaker was in the open position as part of scheduled maintenance. This notification is being made in accordance with 10CFR50.72(b)(3)(iv)(A) due to an actuation that was the result of a valid signal. The 4KV Bus was subsequently restored and the EDG was returned to standby. The licensee notified the NRC Resident Inspector. | |
| ENS 42904 | 13 October 2006 15:34:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Initiation of Reactor Shutdown Required by Technical Specification 3.1.7, Action C.1 | On October 13, 2006, at 1034 hours, a Reactor Shutdown was initiated in response to Technical Specification 3.1.7, Action C.1. At 2236 hours on October 12, 2006, it was determined that the Standby Liquid Control tank had a pinhole leak. In accordance with ASME Code requirements for Class 2 pressure boundaries, the SLC tank was declared inoperable, rendering the SLC system inoperable. TS Action 3.1.7, Action B.1, was entered at 2236 hours on October 12, 2006 and TS Action 3.1.7, Action C.1, was entered at 0636 hours on October 13, 2006. The SLC tank and system remain available. The pinhole leak does not affect the structural integrity of the tank, and the level, temperature and concentration of the boron solution are being maintained per requirements. All ECCS and EDGs are operable and available and the electrical grid is stable. The licensee informed the NRC Resident Inspector. At 1130 CDT, the licensee received a 72 hour extension in getting to Mode 3 (Hot Shutdown). The licensee is currently holding reactor power at 82% in anticipation of bringing the reactor to 100% full power while in parallel performing the necessary repairs to the Standby Liquid Control tank. | |
| ENS 43124 | 26 January 2007 15:15:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Ventilation Air Conditioning (Crev Ac) System Declared Inoperable | On January 26, 2007 at 0915 hours, the Control Room Emergency Ventilation Air Conditioning (CREV AC) system was declared inoperable due to not meeting the surveillance acceptance criteria for the refrigeration condensing unit (RCU) during QCOS 5750-11 (Control Room Emergency Ventilation System Test - performed every 2 years). As result Technical Specification 3.7.5 Condition "A" was entered (30 day Limiting Condition of Operation). The CREV AC system maintains a habitable control room environment and ensures the operability of components in the control room emergency zone during accident conditions. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) because the CREV system is a single train system, and loss of the CREV AC could impact the plant's ability to mitigate the consequences of an accident. At this time, additional analysis is underway to determine the impact of the failure to meet the acceptance criteria on the CREV AC safety function. The NRC Resident Inspector was notified of this event by the licensee.
Further evaluation performed by Quad Cities Station confirms the CREV AC system would have performed its safety function. A revised surveillance (QCOS 5750-11, Control Room Emergency Ventilation System Test) was re-performed successfully, and the CREV AC system was declared operable on January 28, 2007 at 0911 hours. Based on the subsequent evaluation and surveillance testing, ENS report # 43124 is being retracted. R3DO (Dave Passehl) notified. The NRC Resident Inspector was notified of this retraction by the licensee. | |
| ENS 43177 | 20 February 2007 23:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Declared Inoperable Due to an Apparent Equipment Malfunction | On February 20, 2007 at 1720 hours, the High Pressure Coolant Injection (HPCI) system was declared inoperable. During a system walkdown the HPCI Motor Speed Changer (MSC) was discovered energized. The MSC controls HPCI turbine speed during system startup and shutdown, and should not have been energized at the time of the walkdown. Additional troubleshooting determined the MSC was not functioning properly, rendering the HPCI system inoperable. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D) because the HPCI system is a single train system and the loss of HPCI could impact the plant's ability to mitigate the consequences of an accident. In accordance with Technical Specification Action 3.5.1.F, the Reactor Core Isolation Cooling (RCIC) system was confirmed operable. Further troubleshooting and engineering evaluations are continuing. The licensee informed the NRC Resident Inspector.
The purpose of this report is to retract the ENS report made on February 20, 2007 at 1720 hours (ENS #43177) under 10CFR50.72(b)(3)(v)(D), a condition that could have prevented fulfillment of a safety function. The initial report was made when the Unit 2 High Pressure Coolant Injection (HPCI) system was declared inoperable following a system walkdown that discovered the HPCI Motor Speed Changer (MSC) was energized. The MSC controls HPCI turbine speed during system initiation, and should not have been energized at the time of the walkdown since the system was not in operation. During troubleshooting the MSC responded slower than expected. Due to this unexpected behavior, it was not certain if HPCI could have met its design basis requirements. However, a subsequent engineering evaluation has determined that at the time of discovery, the HPCI system injection time would have been sufficient to meet its safety function. Repairs to HPCI were completed and the system was declared operable on February 22, 2007 at 0036 hours. The licensee notified the NRC Resident Inspector. Notified the R3DO (Orth). | |
| ENS 43184 | 25 February 2007 01:20:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Reduction in Emergency Preparedness Capability | On February 24, 2007, at 1820 hours, 13 of 51 alert sirens in the Quad Cities Station Emergency Planning Zone were determined to be inoperable. This is considered a major loss of the Quad Cities offsite notification capability. The alert sirens were disabled due to a loss of power. Power was lost due to a major winter storm in the area. Repairs are in progress at the time of this notification. This report is being made due to the reduction in public notification capabilities in accordance with 10CFR50.72(b)(3)(xiii). He licensee notified the NRC Resident Inspector, and will be notifying the State.
This is a follow up notification to report that only 4 sirens remain inoperable with no estimated time of repair due to downed poles and high voltage power lines on the ground. This is also to correct the number of alert sirens available in the Quad Cities Station Emergency Planning Zone. There are 52 sirens not 51 reported in the original report. With 48 of the 52 alert sirens operational, the station no longer meets the reporting criteria. Notified R3DO (R. Lanksbury) | |
| ENS 43198 | 28 February 2007 07:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Reactor Manually Scrammed Due to Decreasing Condenser Vacuum | At 0120 hours on February 28, 2007 the Unit 2 Reactor was manually scrammed due to decreasing condenser vacuum. All control rods fully inserted during the scram. Reactor water level decreased to approximately -10", which resulted in automatic Group II and III isolations as expected. All systems responded properly to the event. Unit 2 remains in Mode 3, maintaining reactor pressure, with reactor water level in the normal level band. The cause of this event is still under investigation. Unit 1 was unaffected by the event and remains at 97% power. This report is being made in accordance with 10CFR50.72 (b)(2)(iv)(B) and 10CFR50.72 (b)(3)(iv)(A). The Main Condenser remains in service removing decay heat via the bypass valves. All ECCS equipment is available, if needed. The licensee will inform the State and has informed the NRC Resident Inspector. | |
| ENS 43720 | 12 October 2007 21:04:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Made Due to Elevated Levels of Tritium | This notification is being made pursuant to 10 CFR 50.72(b)(2)(xi) due to the fact that Exelon Nuclear has informed the Illinois Emergency Management Agency and Illinois Environmental Protection Agency of recent groundwater monitoring results at Quad Cities Station. This notification was made at 16:04 hours CDST on October 12, 2007. As part of the Station's continuing environmental monitoring and sampling program, new wells were recently installed on Station property to better characterize the known, on-site tritium plume previously identified during Exelon's fleetwide tritium study conducted in 2006. Samples from some of the new wells have indicated elevated levels of tritium requiring notification of the State of Illinois. Based on well data to date, the tritium in the groundwater is confined to Exelon property and poses no threat to public health or safety. The Station continues to track this issue by monitoring the existing wells, installing new wells for further assessment, and pursuing underground pipe integrity testing. The NRC Resident Inspector was notified of this event by the licensee. | |
| ENS 43793 | 20 November 2007 11:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Ventilation System Inoperable | On November 20, 2007, at 0505 hours, the Control Room Emergency Ventilation (CREV) system was declared inoperable due to an inoperable Air Filtration Unit (AFU). During monthly testing it was discovered that the AFU heater failed to operate as required. Technical Specification 3.7.4, Condition A, was entered which requires the CREV system to be restored to an operable status in seven days. This notification is being made in accordance with I0CFR50.72(b)(3)(v)(D), Event or Condition That Could Have Prevented Fulfillment of a Safety Function, because the CREV system is a single train system required to mitigate the consequences of an accident. Troubleshooting is in progress to determine the cause of the AFU heater failure to start. The licensee notified the NRC Resident Inspector and the Illinois Emergency Management Agency (IEMA). | 05000254/LER-2007-003 |
| ENS 44441 | 27 August 2008 14:50:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification and Press Release Due to Inadvertent Airing of a Emergency Broadcast Message | On August 27, 2008 at 0950 CDT, a local radio station (in Clinton, Iowa) mistakenly activated an emergency broadcast message concerning the Quad Cities Station. The inadvertent broadcast message occurred during the radio station's participation in a scheduled Quad Cities Station emergency planning exercise. The radio station has announced a correction to the message and rebroadcasted the correct information. Exelon has notified county and state officials and issued a news release. This event is being reported under 10 CFR 50.72(b)(2)(xi). The licensee has notified the NRC Resident Inspector. | |
| ENS 44602 | 27 October 2008 08:03:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Explosion in the Protected Area | Quad Cities Station has declared an Unusual Event under EAL HU6 due to an explosion in the protected area. The radwaste floor drain surge tank building has physical damage after reports of an explosion in the area. The building is located outside the vital area West of the Turbine building. There are no reports of fire in the area. First responders are on the scene and assessing the situation. This event is being reported under 10 CFR 50.72.(a). The licensee stated that there were no injuries or release of radioactive material and no impact on the reactors. The damage was to block structure of the radwaste floor drain surge tank building but the radwaste tank itself was not impacted. The cause of the explosion is still unknown but appears to be related to a buildup of some kind of hydrocarbon gas. No offsite assistance from the fire department has been requested by the licensee. The explosion was reported to occur at 0242 CDT. The States of Illinois and Iowa have been notified. The licensee has also notified local authorities and will notify the NRC Resident Inspector.
Unusual Event - Termination This is an update to EN #44602. On October 27, 2008, at 0303 hours CDT, Quad Cities declared an Unusual Event under EAL HU6 due to an explosion in the protected area. The radwaste floor drain surge tank building incurred physical damage. A walkdown of the area confirmed that the explosion had no impact on any safety related structures or components. There were no injuries. The combustible atmosphere in the surge tank building has been attributed to the accumulation of methane gas, which originated in the surge tank. Actions have been initiated to purge and monitor the surge tank atmosphere to ensure conditions will not deteriorate. Conditions no longer meet an emergency action level, the event was terminated at 1530 CDT hours on October 28, 2008. A press release will be issued on the termination of the Unusual Event. The licensee has notified the NRC Resident Inspector, Illinois EMA, Illinois REAC, Iowa EMD, Scott County - Iowa, and Clinton County - Iowa. Notified R3RA (Caldwell), R3DO (Kozak), NRR EO (Skeen), IRD (Gott). DHS (Hill), and FEMA (Blankenship). | |
| ENS 45028 | 28 April 2009 23:34:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Process Computer Replacement Impacting Erds and Spds | During the current Unit 1 refueling outage, the plant process computer system will be taken out of service for approximately three weeks to support a planned modification which will install a new (upgraded) plant process computer system. During this timeframe, ERDS (Emergency Response Data System) and SPDS (Safety Parameter Display System) will be unavailable on Unit 1 which will remain in a shutdown condition. The computer outage will also render ERDS and SPDS unavailable on Unit 2 for approximately eight hours. Should the need arise, plant status information will be communicated to the NRC, (State, and local responders) using other available communication systems. The computer outage is expected to commence at 2200 CDT on 4/28/2009, and the process computer will be restored prior to Unit 1 startup currently scheduled for 05/21/2009. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Assessment Capability. The licensee will update this report when ERDS and SPDS is restored to Unit 2. The licensee has notified the NRC Resident Inspector.
Follow-up notification - U2 ERDS and SPDS is restored as of 0315 CDT 4/29/09. The licensee notified the NRC Resident Inspector. Notified R3DO(Hills)
Follow-up notification - U1 ERDS and SPDS have been restored. Acceptance testing was completed on 5/15/09. The licensee notified the NRC Resident Inspector. Notified R3DO(Skokowski) | |
| ENS 45080 | 19 May 2009 14:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Response Data System (Erds) Due to Network Service Not Functioning | On May 19, 2009, it was identified that ERDS was unavailable for approximately 65 minutes (0900 - 1005 hours CDT). It was determined that a network service responsible for broadcasting ERDS data was not functioning. ERDS was restored when the network service was restarted. The cause of the service interruption is under investigation. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a major loss of communications capability. The NRC Resident Inspector has been notified. The State of Illinois will be notified. | |
| ENS 45092 | 25 May 2009 21:50:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Inoperable Rcic Primary Containment Isolation Valves | The Reactor Core Isolation Cooling (RCIC) system tripped on high turbine exhaust discharge pressure during surveillance testing on May 24, 2009. Following the surveillance it was identified that the containment (torus) inboard stop check valve in the RCIC turbine exhaust line could not be closed, necessitating the shutdown of Unit 1 to perform repairs on this inboard isolation valve. At 1650 hours on May 25, 2009, with the unit shutdown, inspections revealed that the inboard isolation valve could not close due to (the presence of) valve material originating from the containment (torus) outboard swing check isolation valve in the turbine exhaust line. Given the inspection findings it has been concluded that neither the inboard stop check valve or the outboard swing check valve in the RCIC turbine exhaust line would have performed their primary containment isolation function while the unit was at power, (thus) impacting containment integrity. As a result, this event is being reported in accordance with 10 CFR 50.72 (b)(3)(ii) for a degraded or unanalyzed condition. The inboard and outboard containment isolation valves in the RCIC turbine exhaust line will be repaired prior to unit startup. The NRC Resident Inspector has been notified of this condition. | |
| ENS 45146 | 21 June 2009 08:08:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Unit 2 Emergency Response Data System (Erds) Due to a Plant Process Computer Issue | On June 21, 2009, it was identified that Unit 2 ERDS was unavailable for approximately 3 hours and 1 minute (0208-0509 hours CDT). This was due to the loss of the Unit 2 Plant Process Computer (PPC) interface with ERDS. Unit 2 ERDS was restored when the Unit 2 PPC interface with ERDS was restored. The cause of the Unit 2 PPC interface interruption is under investigation. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a major loss of communications capability. The Duty NRC Resident Inspector has been notified. | |
| ENS 45269 | 17 August 2009 14:31:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Supervisor Tested Positive for Alcohol | A licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's unescorted access to the plant has been denied. Contact the Headquarters Operations Officer for additional details. The licensee has notified the NRC Resident Inspector. | |
| ENS 45285 | 24 August 2009 07:40:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Tsc Air Filtration System Inoperable Due to Planned Maintenance | Planned maintenance activities will commence today (August 24, 2009) on the Quad Cities Station Technical Support Center (TSC) ventilation air filtration system. The maintenance will be completed in approximately 36 hours, and is scheduled to be worked continuously to minimize out-of-service time. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing procedures; contingency plans are in place to expeditiously restore the TSC ventilation to an operable status. If TSC evacuation is warranted due to radiological conditions, the facility will be relocated in accordance with existing procedures. This event is reportable per 10CFR50.72(b)(3)(xiii) since the scheduled maintenance affects an emergency response facility. The NRC Resident Inspector has been notified.
The TSC air filtration system maintenance has been completed. The TSC has been returned to normal operation. Notified R3DO (Orth). | |
| ENS 45333 | 9 September 2009 01:10:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Reactor Shutdown Required by Technical Specifications Due to an Unisolable Leak in a Core Spray Minimum Flow Line | On September 8, 2009, during performance of quarterly Core Spray (CS) system flow rate testing, a pinhole leak was identified in the 1B Core Spray Minimum Flow Line. The pinhole leak is located just downstream of the 1B CS minimum flow valve (1-1402-38B) and cannot be isolated from primary containment. At 1935, hours it was determined that structural integrity of the piping could not be assured. Technical Specification 3.6.1.1, Required Action B.1, was entered and a plant shutdown was initiated at 2010 hours. In addition, the 1B Core Spray system was declared inoperable. All other ECCS systems remain operable. The NRC Resident Inspector has been notified. This event is being reported under 50.72(b)(2)(i) due to the initiation of a shutdown required by the plant's Technical Specifications. The licensee estimates reactor shutdown will be achieved at 0130 CDT. | |
| ENS 45488 | 11 November 2009 18:38:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Degraded Tsc Emergency Ventilation | At 1238 hours on November 11, 2009, during performance of the Technical Support Center (TSC) ventilation pressure test, it was determined that the TSC could not be maintained at a positive pressure in the emergency mode due to the failure of an outside air intake damper to fully close. At this time, troubleshooting is being performed on the damper. If an emergency is declared requiring TSC activation, the TSC will be staffed and activated using existing procedures. In the event TSC evacuation is warranted due to radiological conditions, key Emergency Response Organization members would be relocated in accordance with existing procedures. This event is reportable per 10 CFR 50.72 (b)(3)(xiii) given the impact on the emergency response facility. The NRC Resident Inspector has been notified. | |
| ENS 45555 | 9 December 2009 17:17:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Winter Storm Power Outages Cause Alert Siren Inoperability | On December 9, 2009, at 1117 hours, 27% of the alert sirens in the Quad Cities Station Emergency Planning Zone were determined to be inoperable for greater than 60 minutes. This is considered a major loss of the Quad Cities off site notification capability. The alert sirens were disabled due to power outages caused by a winter storm. Efforts are underway at the time of this notification to restore the sirens. This report is being made due to the reduction in public notification capabilities in accordance with 10CFR50.72(b)(3)(xiii). A follow-up notification will be provided when the sirens have been restored. The licensee notified the NRC Resident Inspector.
Follow-up Notification - Repairs to the emergency sirens were completed as of 0100 hours. All sirens are restored. The licensee notified the NRC Resident Inspector. Notified R3DO (Phillips). | |
| ENS 45587 | 23 December 2009 12:56:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Winter Storm Power Outages Cause Alert Siren Inoperability | On December 23, 2009, at 0656 hours, 40% of the alert sirens in the Quad Cities Station Emergency Planning Zone were determined to be inoperable for greater than 60 minutes. This is considered a major loss of the Quad Cities offsite notification capability. The alert sirens were disabled due to power outages caused by a winter storm. Efforts are underway at the time of this notification to restore the sirens. This report is being made due to the reduction in public notification capabilities in accordance with 10CFR50.72(b)(3)(xiii). As of 1100 hours, all but 6 EPZ sirens (12%) have been restored. A follow-up notification will be provided when the sirens have been restored.
Repairs to the emergency sirens were completed as of 1456 CST hours. All sirens are restored. | |
| ENS 45771 | 17 March 2010 16:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Erds and Spds Out of Service Due to Process Computer Upgrade | During the current Unit 2 refueling outage, the Unit 2 plant process computer system will be taken out of service for approximately three weeks to support a planned modification which will install a new (upgraded) plant process computer system. During this timeframe, ERDS (Emergency Response Data System) and SPDS (Safety Parameter Display System) will be unavailable on Unit 2 which will remain in a shutdown condition. The computer outage will have no impact on Unit 1. Should the need arise, plant status information will be communicated to the NRC using other available communication systems. The computer outage is expected to commence at 1130 CDT on 3/17/2010 and the process computer will be restored prior to Unit 2 startup, currently scheduled for 4/11/2010. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Assessment Capability. The licensee has notified the NRC Resident Inspector. * * * UPDATE FROM KENNETH OHR TO CHUCK TEAL AT 1410 ON 4/9/10 * * * U2 (Unit 2) ERDS and SPDS have been restored. Acceptance testing was completed on 4/8/2010. Notified R3DO (Peterson). | |
| ENS 45935 | 19 May 2010 18:30:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Potentially Toxic Atmosphere in an Access Faciliity | At 1315 hours personnel performing repairs to an air conditioning system inside the Access Facility (AF) reported a Freon leak from a valve packing failure. All personnel were evacuated from the building. Radiation protection personnel entered the building and atmospheric sampling commenced. Sample results indicated a potentially toxic atmosphere in portions of the AF, and at 1330 hours an Emergency Plan Unusual Event was declared on EAL HU7, 'Hazards and Other Conditions Affecting Plant Security.' The AF was locked down and plant access and egress have been halted pending the restoration of AF atmosphere. Required security compensatory measures are in place. The Freon completely discharged from the affected air conditioner through the leak. Actions to establish a habitable atmosphere inside the AF continue. Units 1 and 2 are unaffected by this event and continue to operate at full power. The licensee notified the NRC Resident Inspector. * * * UPDATE FROM MARK BRIDGES TO PETE SNYDER AT 1551 ON 5/19/2010 * * * A habitable atmosphere has been re-established and conditions no longer meet an emergency action level. The Unusual Event was terminated at 1438 CDT on 5/19/2010. The licensee will notify the NRC Resident Inspector. Notified R3DO (Cameron), IRD (Gott), NRR (Ruland), DHS (Vestal), and FEMA (Barden). | |
| ENS 46169 | 12 August 2010 08:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Reactor Automatic Scram Due to Turbine Trip | While performing a main condenser flow reversal at 0358 (CDT) on August 12, 2010 the Unit 1 reactor received an automatic scram on turbine stop valve closure. Turbine stop valve closure was initiated from a turbine trip on low vacuum. Reactor water level decreased to approximately -3 (inches), which resulted in automatic Group II and III isolations as expected. All systems responded properly to the event. Unit One remains in Mode 3, maintaining reactor pressure with reactor water level in the normal level band. The cause of the event is still under investigation. Unit 2 was unaffected by the event and remains at 100% power. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). All rods fully inserted. Normal feedwater is supplying the reactor with decay heat removal via the turbine bypass valves. No safety relief valves lifted during the transient. The transient was initiated during the condenser flow reversal when one of the flow reversal valves failed to fully reverse the flow. This resulted in inadequate heat transfer to circ water at full power operation causing the condenser vacuum to increase. In addition, the condenser vacuum trip of the turbine appears to have initiated before it was actually required. Overall, the scram was characterized as uncomplicated and all systems functioned as required. | |
| ENS 46184 | 17 August 2010 19:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Initiated Due to Recirculation Pump Trip | On August 17, 2010, at 1414 hours, a manual scram was inserted on Unit 2 due to rising reactor water level. The water level increase was caused by a trip of the 2B Recirculation Pump. Restoration of the 2B Recirculation Pump Programmable Logic Controller (PLC) on the 2B Adjustable Speed Drive (ASD) was in progress at the time of the event. Following the reactor scram, reactor water level decreased to approximately minus 15 inches, which resulted in automatic Group II and III isolations as expected. All systems responded properly to the event. The cause of the event is still under investigation. Unit I was unaffected by the event and remains at 100% power. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). All control rods inserted on the scram. No ECCS initiations occurred. No safety or relief valves lifted. The unit is stable in Mode 3 at 926 psig and 530 degrees Fahrenheit. Pressure is being controlled on the turbine bypass system. The licensee notified the NRC Resident Inspector and the Illinois Department of Nuclear Safety. | |
| ENS 46202 | 24 August 2010 14:22:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Offsite Communications | At 0922 hours (CDT) on August 24, 2010, it was identified that offsite telecommunications capability was lost (including ENS, HPN, and ERDS) for Quad Cities Station when a fiber optic phone cable was cut during offsite work activities. A commercial communication link with the NRC Operations Center was re-established at 1105 hours when site offsite telecommunications were re-routed through the Exelon network to an available commercial system. A follow-up notification will be provided when normal offsite telecommunications have been reestablished. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a major loss of communications capability. The NRC Resident Inspector has been notified.
ERDS has been restored as of 1815 CDT. Notified R3DO(Riemer) | |
| ENS 46535 | 12 January 2011 16:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Loss of Unit 2 Essential Service 480V Bus | On January 12, 2011, at 1020 hours, Unit 2 480V Essential Service Bus 29 was inadvertently deenergized. The cause of the bus trip was inadvertent contact with a bus feed breaker by a station employee during unrelated work activities. Normal power to Bus 29 was restored at 1026 hours. While Bus 29 was deenergized Division II core and containment cooling systems were unavailable and inoperable. The plant responded normally to the loss of Bus 29 with the exception of Bus 28/29-5 which should have automatically transferred from Bus 29 to the reserve feed from Bus 28. This condition renders both divisions of the Low Pressure Cooling Injection (LPCI) mode of the Residual Heat Removal (RHR) system inoperable. Technical Specifications 3.5.1.E has been entered, requiring restoration of LPCI in 72 hours. Bus 28/29-5 was manually energized from Bus 28 at 1213 hours; however, LPCI remains inoperable pending investigation and restoration of Bus 28/29-5 auto-transfer function. This event is reportable under 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented fulfillment of a safety function. LPCI mode of the RHR system is inoperable because the auto-transfer function for Bus 28/29-5 provides a back-up power supply to the LPCI injection motor operated valves from Bus 28. A swing diesel test was in progress during this bus loss, however the diesel was not loaded and had no impact on the bus failure. The diesel was subsequently secured. The licensee notified the NRC Resident Inspector. | 05000265/LER-2011-001 |
| ENS 46769 | 19 April 2011 19:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Divisions of Residual Heat Removal Inoperable | At 1430 hours, a water leak was discovered on the Unit 1 Emergency Diesel Generator Cooling Water Pump (EDGCWP) room cooler. As a result, the Unit 1 EDG (Division II) was declared inoperable and Technical Specification (TS) 3.8.1 was entered. At this time, Division I RHR was inoperable for planned maintenance on the Division I RHR room cooler. In the event of a LOOP/LOCA scenario, RHR would not have been able to perform its safety function (i.e., Division I RHR was impacted by the ongoing room cooler maintenance and Division II RHR would not have had an operable on-site emergency power source). Given the impact on the RHR system, this event is reportable under 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function required to mitigate the consequences of an accident. Note that at the time of discovery, offsite power was available to Division II RHR as well as the Division I and Division II Core Spray systems. At 1816 hours, the Division I RHR system room cooler was restored and Division I RHR was declared operable. The licensee has notified the NRC Resident Inspector. | 05000254/LER-2011-001 |
| ENS 46823 | 5 May 2011 15:35:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center (Tsc) Ventilation System Degraded | At 1035 hours CDT on May 5, 2011, the TSC ventilation was secured due to the performance degradation of the air handling unit's fan. The air handling unit is required to maintain a positive pressure in the TSC when the ventilation system is in the emergency mode. At this time, repair activities are being pursued for the fan. If an emergency is declared requiring TSC activation, the TSC will be staffed and activated using existing procedures. In the event TSC evacuation is warranted due to radiological conditions, key emergency response organization members would be relocated in accordance with existing procedures. This event is reportable per 10 CFR 50.72 (b)(3)(xiii) given the impact on the emergency response facility. The NRC Resident Inspector has been notified. | |
| ENS 46951 | 13 June 2011 10:10:00 | 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 Steam Leak in Turbine Building | On June 13, 2011 at 0510 CDT, a manual scram was inserted on Unit 1 following discovery of a steam leak in the turbine building downstream of the Main Turbine Stop Valves. Following the reactor scram, reactor water level decreased to approximately -19 inches, which resulted in automatic Group II and Group III isolations (expected response). The steam leak was isolated by closure of the Main Turbine Stop Valves. All systems responded properly to the event. Unit 1 remains in Mode 3 (hot shutdown), maintaining reactor pressure with the Main Turbine Bypass Valves; reactor water level is in the normal level band. The cause of the event is still under investigation. Unit 2 was unaffected by the event and remains at 100% power. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). The licensee has notified the NRC Resident Inspector. | |
| ENS 47030 | 6 July 2011 19:10:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Ventilation Ac System Inoperable | On July 6, 2011 at 1410, the Control Room Emergency Ventilation Air Conditioning (CREV AC) system was declared inoperable due to unsatisfactory performance, high vibration condition of the air conditioning compressor. As a result, Technical Specification 3.7.5 condition A was entered. The CREV AC system maintains a habitable control room environment and ensures the operability of components in the control room emergency zone during accident conditions. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) because the CREV system is a single train system, and loss of the CREV AC could impact the plant's ability to mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector. | 05000254/LER-2011-003 |