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| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 40123 | 31 August 2003 06:32:00 | Other Unspec Reqmnt | 24-Hour Condition of License Report Involving Potential Violation of Maximum Power Level | This 24-hour report is being made as required by Braidwood Unit 2 License Condition 2.G as a potential violation of the maximum power level (3586.6 Mwt) as stated in Unit 2 License Condition 2.C(1). As a result of issues at Byron Unit 1 and Unit 2 concerning potential discrepancies in the ultrasonic flow measurements for the main feedwater system, Braidwood investigated both Unit 1 and Unit 2 to determine if similar issues existed. These flow measurements are used in the calorimetric calculation for reactor power. Ultrasonic flow measurements were taken on the four individual main feedwater lines on Braidwood Unit 2. These measurements identified the presence of flow signal noise in the data signals for two of the four ultrasonic flow measurement devices installed on the individual feedwater lines, which may adversely affect the integrity of these measurements. In response to identifying this flow signal noise, Braidwood removed corrections based on ultrasonic flow measurement from these two loops. Based on removing credit for these ultrasonic flow measurements, it was determined at 0132 on August 31, 2003, that Braidwood Unit 2 could have potentially exceeded its licensed thermal power limit by up to 0.8%. Ultrasonic flow measurements were taken on the main feedwater system piping header on Braidwood Unit 1 and were compared to the results from the ultrasonic flow measurement devices on the four individual feedwater lines. Based on the results of the data analysis, Unit 1 was determined to be acceptable. The power level on Unit 2 was reduced to less than 100% power consistent with the feedwater flow as measured directly by the venturis without using the correction factor on two of the four ultrasonic flow meters. Additional actions regarding the investigation of the condition, determination of the root cause and corrective actions, and the determination of the potential actual overpower will be included in the 30-day license event report. The licensee informed the NRC Resident Inspector. HOO NOTE: See Byron Event Notification #40117.
The following information was provided by the licensee: The venturis were installed during initial construction of both Units. They are the original design for the Units. AMAG is a method of calibrating our Feedwater Venturi flow instruments. These devices were installed by the Advanced Measurement and Analysis Group (AMAG). AMAG first applied: Unit 1 - 6/11/1999 and Unit 2 - 6/11/1999 Rated MWt (Mega Watt Thermal) 3411 to 3586.6: Unit 1 - 5/14/2001 (and) Unit 2 -5/24/2001 Of the above 5% uprate to 3586.6 MWt, we initially increased power only 1 % of the above. This 1 % increase was known as our mini-uprate. We did not increase the full 5% since our turbines had not yet been modified to withstand the higher flow rates. During the A_R09 outages, we modified the turbines and achieved full uprate as follows: Full Uprate to 3586.6 MWt: Unit 1 - 10/16/2001 (and) Unit 2 - 5/15/2002. Notified NRR (Reis) and R3DO (Gardner) | |
| ENS 40130 | 3 September 2003 12:30:00 | Other Unspec Reqmnt | 24-Hour Report Required by License Condition -- Potential Violation of Maximum Power Level | The following report was submitted by the licensee via fax: This 24-hour report is being made as required by Braidwood Unit 1 License Condition 2.G and Braidwood Unit ~ License Condition 2.G as a potential violation of the maximum power level (3586.6 MWt) as stated in Unit 1 and Unit 2 License Condition 2.C(1). Braidwood received Nuclear Safety Advisory Letter (NSAL) 03-6 from Westinghouse Electric Company. This NSAL documented that errors were found in calculations that may result in the use of a non conservatively high net heat input value to the plant calorimetric calculation. The net heat input is the difference between the reactor core power and the nuclear steam supply system power. The values used in the NSAL were based on generic values for certain operational parameters. The NSAL documented that an increase in actual reactor power could be as much as 0.4 MWt. This equates to an error in reactor power of approximately 0.011%. Braidwood reviewed the power history and 8-hour average calorimetric values on Unit 1 and Unit 2 since full power uprate was applied on each Unit. Several time periods were identified on each Unit where the 8-hour calorimetric value exceeded the license thermal power limit when the 0.011% error was added to the 8-hour calorimetric value. Therefore, the licensed power limit of 3586.6 MWt was slightly exceeded on both Unit 1 and Unit 2. The power level on both Units was reduced to less than 100% to account for the 0.4 MWt error. The calorimetric program was then updated to account for this error. This issue was identified at 0730 on September 3, 2003, and has been entered into the Corrective Action Program. Additional information will be contained in the 30-day licensee event report. The licensee has notified the NRC Resident Inspector. | |
| ENS 40298 | 5 November 2003 03:32:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Braidwood 2 Afw Support System Actuation During Outage | Unit 2 auxiliary feedwater support systems actuated during scheduled ATWS testing when an unrelated clearance order placement de-energized two 6.9 kv busses (256 and 258). The 2 of 4 6.9 kv bus undervoltage coincidence initiated a valid auto-start signal causing lube oil pumps 2AF01PA-A, 2AF01PB-A and 2AF01PB-C to start. Auxiliary Feedwater Pump AOV discharge valves 2AF004A and 2AF004B auto opened. 2AF01PA 4kv breaker, which was in the equipment test position. Neither auxiliary feedwater pump started and no water transferred to the steam generators. The licensee notified the NRC resident inspector. | |
| ENS 40370 | 3 December 2003 09:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Plant Had an Auto Reactor Trip from 100% Power Due to Steam Generator Low Level | The "2 D" steam generator Lo-2 level was caused by the loss of the "2C" feedwater pump while performing the "2 BWOS" feedwater weekly surveillance of the HP stop valve. Both trains of the aux feed actuated as expected on the "2D" Lo-2 s/g level signal. The plant is currently in mode 3 with all rods fully inserted. No ECCS or safety relief valves actuated. Licensee notified the NRC Resident Inspector | |
| ENS 40559 | 1 March 2004 22:00:00 | Other Unspec Reqmnt | 24-Hour Condition of License Report Involving Potential Violation of Maximum Power Level | This 24-hour report is being made as required by Braidwood Unit 1 License Condition 2.G as a potential violation of the maximum power level 3586.6 MWt as stated in Unit 1 License Condition 2.C(1). Braidwood is conservatively reporting an overpower condition on Unit 1 due to the implementation of an ultrasonic flow measurement system. Unit 1 was potentially overpowered a maximum of 1.07%. This value is based upon the maximum ultrasonic flow meter correction factor used for Unit 1 during the period between June 1999 and September 2003. During post installation testing of the permanent ultrasonic flow measurement system, discrepancies in the ultrasonic flow measurement system were identified that could have resulted in an overpower condition of Unit 2 during the previous use of the ultrasonic system (i.e., during the period between June 1999 and September 2003). The ultrasonic flow measurement system was used to correct venturi feedwater flow measurements. The corrected feedwater flow measurements were then used is the calorimetric calculation for reactor power. Currently Braidwood Unit 1 and Unit 2 are controlling power level based only on the venturi feedwater flow indication. The licensee notified the NRC Resident Inspector. | |
| ENS 40762 | 20 May 2004 19:35:00 | 10 CFR 26.73, Applicability | Corporate Headquarters Non-Licensed Employee Tested Positive During Urinalysis | A non-licensed supervisory employee with normal work location at corporate headquarters tested positive during a random urinalysis test. The employee's access to Braidwood, Byron, Clinton, Dresden, LaSalle and Quad Cities nuclear power stations has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee will notify the NRC Resident Inspectors for the above sites. | |
| ENS 40922 | 4 August 2004 12:15:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Notification Due to Inoperable Sirens | At 0715 on August 4, 2004, Exelon was notified by Fulton Contracting that Braidwood Station had >(greater than) 25% of its emergency notification sirens inoperable. The cause of the inoperability (at 0626) for the sirens was due to a loss of power due to severe weather. As of 0708 on August 4, 2004, the number of inoperable sirens was less than 25%. Due to a major loss of emergency preparedness capabilities, this event is reportable per 10 CFR 50.72(b)(3)(xiii). The licensee notified the NRC Resident Inspector. | |
| ENS 41038 | 13 September 2004 20:46:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Minor Oil Spill | While performing routine ditch cleaning maintenance (on Sept. 13, 2004), a faulty hydraulic hose on a bulldozer caused a small release of hydraulic oil near a drainage ditch adjacent to Braidwood Station property. The hose was intact prior to maintenance beginning on the morning of Sept. 13. It is estimated that 10-15 gallons of oil spilled on the ground, with a very small portion of the oil (less than one gallon) going into the drainage ditch itself. The oil on the ground has been removed. A compensatory oil boom was placed downstream of the spill in the ditch to collect the small amount of oil that went into the ditch. A thorough walk down of the ditch, downstream from the spill, revealed no evidence of oil. The Station will investigate why the hose failed and what actions can be taken in the future to prevent recurrence. Per Station procedures, official notifications were made to the Illinois Emergency Management Agency, and the National Response Center. In addition, Exelon communications notified the Village of Godley, the Godley Park District, the Village of Braceville, and the Braceville Fire Department District. Residents in the area were notified through personal visits. The licensee notified the NRC Resident Inspector. | |
| ENS 41280 | 22 December 2004 19:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Actuation Due to Steam Generator Low Level Signal | Unit 2 reactor trip due to 2C steam generator LO-2 reactor protection signal. Auxiliary Feed Water actuated as expected. No additional malfunctions or unexpected plant response. Cause of LO-2 steam generator reactor protection signal under investigation. This is a 4 hour notification of an RPS actuation per 10 CFR 50.72(b)(2)(iv)(B). The 8 hour notification of Auxiliary Feed Water system actuation per 10 CFR 50.72 (b)(3)(iv)(B) is being made under this same telephone call. The licensee notified the NRC Resident Inspector. All control rods fully inserted. Decay heat is being removed to the main condenser via the turbine by-pass valves. The electrical grid is stable.
The licensee has determined that the RPS and Auxiliary Feed Water actuations were the result of an actual low level in the 2C steam generator. The cause of the low level is under investigation. The NRC Resident Inspector was informed. Notified R3 DO (L. Kozak). | |
| ENS 41534 | 28 March 2005 19:59:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Hydrogen Leak | The licensee faxed the following information: Following the Unit 2 trip, (due to a malfunction of the generator protection circuitry), a hydrogen leak was identified on the Unit 2 main generator. The leakage was sufficient enough to cause a flammable gas release affecting normal plant operation. The Unusual Event was declared under HU6 - Hazards and Other Conditions. The State and local authorities were notified at 14:09 CST. There is NO fire, it is a hydrogen release only. The licensee said the hydrogen totalizer, which indicated a flow rate of 100 cfm, may be a probable area of leakage. The hydrogen leaked directly into the turbine building. Air samples indicated personnel breathing apparatus was not required. The licensee expects to exit from the UE once the hydrogen system was purged with CO2. Estimated timeframe is 3-6 hours. The licensee notified the NRC Resident Inspector.
The licensee terminated their Unusual Event at 16:40 CST, and the plant status is 0% power / Mode 3. The hydrogen leak was determined to be from a bushing on the main generator. The hydrogen leakage is believed to have lasted only 5 minutes. The licensee notified the State and NRC Resident Inspector. Notified R3DO (Riemer), IRD Manager (Crlenjak), NRR EO (Berkow), FEMA and DHS.
Received Event Summary Report from Braidwood station. Corrected Unusual Event termination time from 16:40 to 16:23 CST. Notified R3DO (Riemer) and NRR EO (Brenner) | |
| ENS 41535 | 28 March 2005 18:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Generator Protection Circuitry | The licensee faxed the following: Unit 2 reactor trip due to generator protection circuitry. Auxiliary feedwater actuated as expected. There were no additional malfunctions or unexpected plant response. The cause of the generator protection circuitry induced trip is still under investigation. This is a 4 hour notification of an RPS actuation per 10CFR 50.72(b)(2)(iv)(B). The 8 hour notification of an auxiliary feedwater system actuation per 10CFR 50.72(b)(3)(iv)(A) is being made under the same telephone call. See Event 41534. The licensee notified the NRC Resident Inspector. | |
| ENS 41627 | 23 April 2005 13:30:00 | 10 CFR 26.73, Applicability | Fitness for Duty: Alcohol | A non-licensed contract employee was determined to be under the influence of alcohol during a for cause test. The employee's access to the plant has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee will notify the Resident Inspector. | |
| ENS 41807 | 28 June 2005 18:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to Illinois Department of Natural Resources | A 1300 CDST on 06/28/05, it was discovered that a minor fish kill took place in Braidwood Nuclear Station's cooling lake, due to increased temperature. As Summer air temperatures increased, the cooling lake temperature increased above 95�F, the threshold at which species susceptible to increased temperature begin to perish. It was communicated to the Illinois Department of Natural Resources, that approximately 1000 fish, the majority of which were Gizzard Shad, had perished. This notification is made in accordance with Exelon Reportability manual SAF 1.9, ENV 3.26, Braidwood Station Facility Operating License Appendix B Section 4.1 'Unusual or Important Environmental Events,' and 10CFR50.72(b)(2)(xi). The licensee notified the NRC Resident Inspector. | |
| ENS 41858 | 21 July 2005 00:17:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Emergency Sirens Failed for One Hour | At 2017 CDST on July 20, 2005, it was determined that greater than 25% of Braidwood Station Emergency Sirens, which are maintained by others, had been failed for one hour. The initial failure occurred at 1917 CDST on July 20, 2005, and was apparently caused by storm activity in the area. At 2041 CDST on July 20, 2005, Braidwood Station was notified by the Corporate Emergency Planning organization that the number of failed Emergency sirens was less than 25% of the total number of sirens, and that sirens were in the process of being restored at that time. This notification is required by 10CFR 50.72(b)(3)(xiii) The licensee notified the NRC Resident Inspector. | |
| ENS 42184 | 2 December 2005 16:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification - Elevated Levels of Tritium Found in Groundwater | This notification is being made pursuant to 10 CFR 50.72(b)(2)(xi) for a press release issued by Exelon Nuclear at 10:00 AM CST on December 2, 2005 regarding elevated levels of tritium found in groundwater on the Braidwood Station site property near the plant's north boundary. An environmental monitoring program at the Braidwood Generating Station has found higher than normal concentrations of tritium close to an underground pipe inside the plant's northern boundary, and the station has begun a remediation program. An Exelon Nuclear environmental team is drilling test wells on and just beyond the Braidwood property line in order to determine how much tritium may have moved beyond the plant boundaries and ultimately to clean up the tritium. Exelon Nuclear has notified NRC regional personnel, appropriate state agencies, local and state elected officials and four property owners who are potentially affected. The tritium was found in shallow groundwater 8 to 15 feet deep on company property. It poses no health or safety risk to the public and does not threaten drinking water wells in the area. Tritium is a naturally occurring isotope of hydrogen that emits a very low level of radiation and is a natural part of water. It is found in more concentrated levels in water used in nuclear reactors. The closest private residential wells to the site showed no tritium above natural background levels. A sample of water from a pond 50 yards north of the plant property line showed tritium levels of about 2,400 picocuries per liter, above background levels but less than one-eighth of the federal drinking water limit. The residential and pond test samples were taken with the consent of property owners and the results received on Dec. 1. The underground pipe that passes near the monitored site in the past has carried water containing tritium from the plant to the Kankakee River, where it was periodically discharged under federal guidelines as part of normal plant operations. No tritiated water is currently in the pipe and no tritium is currently being introduced into the ground. Braidwood has not released levels of tritium that exceeded federal limits. The licensee notified the NRC Resident Inspector, State and local agencies and has issued a press release. | |
| ENS 42187 | 2 December 2005 23:00:00 | Other Unspec Reqmnt | Reactor Power Operation in Excess of Operating License Condition | This notification is being made pursuant to Braidwood Station Unit 1, Operating License Condition 2.G, which requires a 24 hour notification to the NRC Operations Center for reactor power operation in excess of 3586.6 megawatts thermal (100 percent rated power), i.e., a violation of Operating License Condition 2.C (1)- Braidwood Station will, following this notification, provide a written report within thirty days in accordance with the procedures described in 10 CFR 50.73(b), (c), and (e). Braidwood Station, Unit 1 experienced a Feedwater temperature transient on November 18, 2004, which caused reactor power to momentarily increase and peak at approximately 101.2% as indicated on the excore nuclear instrumentation system. The duration that reactor power remained above 100% was approximately one minute. Subsequent to the event, a number of peer reviews were conducted to validate that power did not exceed 102%. The results of these reviews questioned the methodology used to determine the power level at the time of the transient. Industry was consulted regarding methodologies appropriate for power level measurement during transient conditions. At 1700 on December 2, 2005, an independent Exelon task force concluded, using a conservative methodology, that reactor power level during this transient did exceed 100% for approximately one minute and was limited to a peak of approximately 103.5% and that the appropriate reports, as specified in the Operating License, should be initiated. This overpower transient, caused by the loss of a Feedwater heater string, is bounded by the Feedwater design basis transient described and analyzed in UFSAR, Section 15.1.1, "Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature, therefore, the event did not place Braidwood Unit 1 in an unanalyzed condition that significantly degraded plant safety. No safety limits were exceeded and there was no impact on the health and safety of the public. The licensee has notified the Resident Inspector. The licensee notified the NRC Resident Inspector. | |
| ENS 42339 | 15 February 2006 19:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification - Press Release Concerning Tritium Inspection Program | This notification is being made pursuant to 10 CFR 50.72(b)(2)(xi) to let the NRC know that a press release issued by Exelon Nuclear at 1300 CST on February 15, 2006 mentions, in part, recently identified inadvertent tritium releases from Braidwood, Byron and Dresden Stations. This event was reported by Exelon Corporate also includes Byron 1 & 2 ( both units at 100%) and Dresden 2 & 3 ( Unit 2 - 97% and Unit 3 - 98%). See related Braidwood Station event - Event Number 42184. The licensee will notify the NRC Resident Inspector. | |
| ENS 42472 | 5 April 2006 11:35:00 | 10 CFR 26.73, Applicability | Fitness for Duty - Confirmed Positive for Non-Licensed Contractor Supervisor | A non-licensed contract supervisor had a confirmed positive for alcohol while attempting to gain access the plant this morning. The contractor's access to the plant has been suspended. A review of previous work is being performed. Contact the Headquarters Operations Officer for additional details. The NRC Resident Inspector has been notified. | |
| ENS 43041 | 30 October 2006 11:37:00 | 10 CFR 50.73(a)(1), Submit an LER | Inadvertent Opening of Two Isolation Valves | This telephone notification to report an invalid actuation is provided in accordance with 10 CFR 50.73(a)(1), which states, 'In the case of an invalid actuation reported under Sec. 50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. The specific reporting requirement in 10 CFR 50.73(a)(2)(iv)(A), states, 'Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B).' For this report, the affected system was the Unit 2 Train B Containment Spray System. On October 30, 2006 at approximately 0600 hours, a Main Control Room (MCR) Senior Reactor Operator identified during a panel walk-down that the containment spray header isolation valve, 2CS007B, and the containment spray eductor NaOH tank suction isolation valve, 2CS019B, were open. This was an unexpected condition, which prompted Operations to investigate. A review of the Alarm computer points revealed that the valves opened at 0537 hours. The investigation determined that the most probable cause of this occurrence was attributed to, during the performance of maintenance inside of the 2PA10J panel, an Electrical Maintenance Department technician inadvertently making contact with the manual latch causing the actuation of the K643B slave relay that led to repositioning of the out-of-position valves. The following information provides the required details outlined in NUREG 1022 Revision 2: (a) The relay actuation for the 2B Containment Spray valves was not a valid ESF actuation. The actuation was the result of an inadvertent bump of the K643B relay and not the result of a valid ESF signal from the reactor protection system. (b) This report is being made under 10CFR50.73(a)(2)(iv)(A). (c) The specific train and system that actuated was 2B Containment Spray. (d) The train actuation was a partial actuation for three containment spray valves from the K643B slave relay due to an inadvertent bump of the slave relay manual latching mechanism with the following conditions identified: " 2CS007B, 2B containment spray header isolation valve, changed from closed to open. "2CS010B, 2B containment spray NaOH eductor inlet isolation valve, is normally open and did not change position, but received an open signal. "2CS019B, 2B containment spray eductor NaOH tank suction isolation valve, changed from closed to open. The 2B containment spray system was in TEST in preparation for changing from Mode 5 to Mode 4. (e) The containment spray system did not start because the system was in TEST and the containment spray pumps were in lock out while in Mode 5. The three valves functioned as designed. The licensee notified the NRC Resident Inspector. | |
| ENS 43131 | 29 January 2007 17:30:00 | 10 CFR 26.73, Applicability | Licensed Employee Fails Fitness-For-Duty Test | A licensed operator employee had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | |
| ENS 43137 | 1 February 2007 15:25:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Offsite Phone Connections | At approximately 0925 CST, Braidwood lost approximately 85% of phone functionality. The loss of phones has been determined to be an off site issue and the local phone company is investigating the situation. The Emergency Response Organization (ERO) can still activate ERO pagers using cell phones. The Shift Manager has a dedicated cell phone. The licensee notified the NRC Resident Inspector.
The licensee declared the ENS and commercial lines operable based on testing and information from the telephone company. The R3DO (Hills) has been notified. | |
| ENS 43449 | 27 June 2007 14:21:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Off-Site Power Fluctuation | Switchyard line 2001 tripped and re-energized during a thunderstorm. This caused main generator output breaker, ACB 3-4, to trip open. At this time '1D' reactor coolant pump (RCP) tripped and caused a reactor trip. Cause for the '1D' RCP trip is under investigation. The heat sink is being provided by Aux Feedwater and the use of Steam Dumps. Electrical power is being provided by offsite power. All rods inserted. There were no complications during the trip and all systems functioned as required. Offsite power remained available throughout the transient. The electrical transient had no impact on Unit 2. The licensee notified the NRC Resident Inspector.
This is a revision to a previously transmitted ENS call on 6/27/07 EN# 43449. Braidwood Unit 1 Reactor automatically tripped on a loss of '1D' RCP greater than P-8 setpoint. The cause of the RCP trip was an electrical disturbance during a thunderstorm. The reactor trip automatically caused a main turbine and generator trip. Auxiliary feedwater system automatically started on the low-2 S/G water level that is expected from a full power reactor trip. Auxiliary feedwater and main steam dumps are providing a heat sink. A switchyard line also tripped during this electrical transient causing multiple switchyard breakers to open. Electrical power is being provided by offsite power. All control rods fully inserted. There were no complications during the trip and all systems functioned as required. Offsite power remained available throughout the transient. The electrical disturbance had no impact on Unit 2. The licensee notified the NRC resident inspector." R3DO (Louden) notified. | |
| ENS 43590 | 23 August 2007 20:30:00 | 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)(3)(xiii), Loss of Emergency Preparedness | Manual Reactor Trip Because of Lowering Condenser Vacuum | At 1530 hours on 8/23/07, Braidwood Station Unit 2 was manually tripped due to lowering condenser vacuum. The lowering condenser vacuum resulted from the trip of two circulating water pumps. The cause of the two circulating water pump tripping is under investigation. All control rods inserted and there were no complications during the trip and all systems functions as required. Following the unit trip, the Auxiliary Feedwater System actuated as expected to maintain steam generator level. At the time of the unit trip, the Braidwood Station area was experiencing severe thunderstorms. Additionally, at 1604 hours, 19 of 70 emergency sirens for the Braidwood Station were declared inoperable due to a loss of power from storms in the area. As of 1704 hours, 19 sirens (greater than 25%) remain inoperable. This event is considered a major loss of offsite response capability and applies to both Braidwood Station Unit 1 and Unit 2. These events are is being reported under: (1) 10 CFR 50.72(b)(2)(iv)(B) as an event that results in the actuation of the reactor protection system (RPS) when the reactor is critical, (2) 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PWR auxiliary feedwater system. (3) 10 CFR 50.72(b)(3)(xiii) as a major loss of offsite response capability. All safety buses remained powered by offsite power throughout this event. Emergency diesel generators are available if needed. No steam generator PORV's lifted as a result of the trip. Decay heat is being discharged to the condenser via the steam dumps. The licensee informed the NRC Resident Inspector. | |
| ENS 44354 | 17 July 2008 16:41:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Degraded Control Room Ventilation Envelope | At 1141 on Thursday, July 17, 2008, Braidwood Station determined that the Control Room Envelope (CRE) boundary had a degraded condition based on data obtained during performance of differential pressure testing performed in accordance with Technical Specification (TS) 5.5.18, 'Control Room Envelope Habitability Program.' Degradation of the CRE boundary was identified that could have prevented the Control Room (VC) Ventilation Filtration System from performing its safety function due to the potential for a greater amount of unfiltered inleakage into the CRE than assumed in the licensing basis analysis for Design Basis Accident consequences. The differential pressure test results identified slightly negative pressures as compared to outside air in some areas of the CRE that do not include spaces that control room occupants inhabit during accident conditions. In accordance with the requirements of TS 3.7.10, 'VC Filtration System,' Condition B, mitigating actions have been implemented to lessen the effect on CRE occupants from the potential hazards of a radiological or chemical event or a challenge from smoke. Actions are in progress to resolve the CRE degraded condition. This event is being reported under 10CFR50.72(b)(3)(v)(D) as an 8-hour report as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident. A Licensee Event Report will be submitted under 10CFR50.73(a)(2)(v). The licensee informed the NRC resident.
The licensee is retracting this report based on the following: On July 17, 2008, NRC Notification 44354 was conservatively made pursuant to 10 CFR 50.72(b)(3)(v)(D) related to the failure of a control room envelope (CRE) differential pressure test. This was the first time this test methodology had been used at Braidwood Station. Specifically, based on the test results, the unfiltered CRE inleakage could not be confirmed as being bounded by the values assumed within the accident dose analyses and hence the Control Room Ventilation (VC) Filtration System could have been prevented from performing its specified safety function. Subsequent evaluation has determined that the VC Filtration System was capable of performing its specified safety function to maintain CRE habitability, since the identified condition would not have resulted in a greater amount of unfiltered inleakage into the CRE than assumed in the licensing bases analyses. Therefore, this event notification is being retracted. The causes for the test failure were attributable to an incorrect test methodology and minor CRE boundary imperfections that had existed since plant construction. These CRE minor imperfections were present during performance of the last CRE unfiltered air inleakage test performed in 2004, which produced acceptable results. Both of these causes were corrected and the test reperformed with acceptable results. Evaluation of this event notification is documented in the corrective action program. The NRC resident has been notified of this retraction. Notified R3DO (Passehl). | |
| ENS 44743 | 27 December 2008 20:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip as a Result of a Generator Trip | At 1418 on 12-27-08 Braidwood Unit 2 experienced an automatic Reactor Trip. The Reactor Trip red first out annunciator was Turb(ine) Trip above P8 Rx Trip. At the time of the trip the Unit Aux Transformer (UAT) 241-1 sudden pressure relay actuated causing a main generator trip which resulted in a main turbine trip which resulted in a Reactor Trip. Also at the same time as the Reactor Trip, the 2C Heater Drain Pump tripped on phase A over current. Damage was subsequently noted on the pump motor terminal box. No fire or smoke was observed at UAT 241-1 or the 2C Heater Drain Pump. After the Reactor Trip occurred, all four steam generators reached their low-2 Reactor Trip setpoints and the pressurizer reached its low pressure Reactor Trip setpoint all of which is an expected response on a trip from full power. Steam generator levels and pressurizer pressure have been restored. Both the 2A and the 2B Auxiliary Feedwater Pumps auto started on the low-2 steam generator levels as expected. All control rods fully inserted into the core. No secondary relief valves lifted and no secondary steam was released as a result of the Reactor Trip. Steam generators are now being filled by the Startup Feedwater Pump and the Auxiliary Feedwater Pumps have been placed in standby. The main steam dumps are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. This report is being made per 10CFR 50.72(b)(2)(iv)(B) for RPS actuation, 4 hr (notification), and per 10CFR 50.72(b)(3)(iv)(A) for automatic actuation of the Auxiliary Feedwater System, 8 hr (notification). The electrical line up transferred to the normal shutdown configuration with standby diesel generators and safety systems available. There was no impact on Unit 1. The licensee plans on issuing a press release and has notified the NRC Resident Inspector. | |
| ENS 45017 | 24 April 2009 16:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Instrument Calibration | At 1141 CT, Braidwood Unit 2 experienced an automatic Reactor Trip. The Reactor Trip red first out annunciator was Over Temperature Delta Temperature (OTDT). At the time of the Reactor Trip the Instrument Maintenance Department was performing a scheduled calibration of a Pressurizer Pressure channel (2PT-456) which is in the B loop of reactor protection. During the calibration a spike occurred on the D loop of reactor protection. Specifically, the RCS (Reactor Coolant System) temperature for the D loop. This caused a Reactor Trip on a 2 of 4 coincidence. After the reactor trip occurred, all four steam generators reached their low-2 Reactor Trip setpoints and pressurizer pressure reached its low pressure Reactor Trip setpoint all of which is an expected response on a trip from full power. Steam Generator levels and Pressurizer pressure have been restored. Both the 2A and 2B Auxiliary Feedwater pumps auto started on the low-2 steam generator levels as expected. All control rods fully inserted into the core. No secondary relief valves lifted and no secondary steam released as a result of the Reactor Trip. Steam Generators are now being filled by the 2A Main Feedwater pump and the Auxiliary Feedwater pumps have been placed in standby. The main steam dumps are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation, 4 hour notification, and per 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the Auxiliary Feedwater System, 8 hour notification. The electrical line up transferred to the normal shutdown configuration with the standby diesel generators and safety systems available. There is no Unit 1 impact. The licensee plans on issuing a press release and has notified the NRC Resident Inspector. | |
| ENS 45104 | 29 May 2009 02:40:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Loss of Control Power to Eccs Valves | At 2140 on 5/28/09 Braidwood Station identified a loss of control power for a Safety Related MCC (Motor Control Center) which provided power to SVAG (Single Valve Actuation Group) valves associated with both trains of the ECCS system. The MCC is normally de-energized to maintain the valve power removed in accordance with Tech Specs for ECCS. Loss of the control power for the associated MCC would prevent operation of these valves, which would prevent realignment of components required for transfer to cold leg recirculation and hot leg recirculation for long term core cooling. Entry was made into LCO 3.5.2, ECCS Operating, and LCO 3.0.3 due to inoperability of both trains of ECCS based on the inability to realign portions of both trains of the ECCS system from injection to cold leg recirculation and subsequent hot leg recirculation. At 2230 on 5/28/09 preparations had been completed for ramp off line per LCO 3.0.3. Troubleshooting was performed and a blown control power fuse was identified and replaced at 2319 on 5/28/09. No Unit ramp was initiated. The NRC Resident Inspector was notified.
The purpose of this report is to retract the ENS report made on May 29, 2009 at 04:49 EDT (ENS #45104) under 10CFR50.72(b)(3)(v)(B), a condition that could have prevented fulfillment of a safety function. The initial report was made based on identification of a loss of control power for a Safety Related MCC (Motor Control Center), which provided power to SVAG (Single Valve Actuation Group) valves associated with both trains of the ECCS system. The MCC is normally de-energized to maintain the valve power removed in accordance with Technical Specifications. The loss of the control power for the associated MCC would prevent operation of these valves. It was initially concluded that this condition would prevent realignment of components required for transfer to cold leg recirculation and hot leg recirculation for long term core cooling. Therefore, the referenced ENS report was made for a loss of safety function. Subsequent review of UFSAR information and previously developed analytical data determined that the safety function for ECCS was not lost due to the event. The failure of the MCC to energize would have NOT affected the ability of the 1B ECCS train to perform its design function of cold and hot leg recirculation. A blown control power fuse, the cause of the event, was identified and replaced on May 28, 2009 at 23:19 hours. The NRC Senior Resident Inspector has been notified of this retraction. R3DO(Pelke) notified. | |
| ENS 45161 | 24 June 2009 21:39:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Illinois Department of Natural Resources Notified of a Fish Kill | At 1639 CDT on 6/24/09, it was discovered that a fish kill took place in the Braidwood Station's cooling lake due to increased temperature. As summer air temperature increased, the cooling lake temperature increased above 93 degrees F, the threshold at which species susceptible to increased temperature begin to perish. It was communicated to the Illinois Department of Natural Resources (IDNR) that a fish kill has occurred. This notification is being made in accordance with the Exelon Reportability Manual section SAF 1.9, Braidwood Station Facility Operating License Appendix B Section 4.1,'Unusual or Important Environmental Events', and 10CFR50.72(b)(2)(xi). See also EN #45151 for similar event. The licensee notified the NRC Resident Inspector. | |
| ENS 45238 | 31 July 2009 02:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to a Loss of Offsite Power for Greater than 15 Minutes | Unit 2 automatically tripped from 100% reactor power as a result of the over-current trip of the 2C Reactor Coolant Pump. Both station auxiliary transformers on Unit 2 subsequently tripped offline. All control rods fully inserted on the trip. Auxiliary feedwater auto-started and maintained Steam Generator water level. The unit is stable in Mode 3. The Emergency Diesel Generators auto started and loaded supplying both emergency busses with power. All systems functioned as required. There was no affect on Unit 1. The licensee notified the NRC Resident Inspector.
At 2059 on July 30, 2009, a reactor trip of Unit 2 at Braidwood occurred. A loss of offsite power occurred and an Unusual Event was declared at 2108. NRC Headquarter Operations was notified at 2155 (ENS call # 45238). Power from System Auxiliary Transformer (SAT) (credited offsite power supply) 242-2 was restored to buses 241 and 242 (safety related buses) at 0036 on August 2, 2009. The Unusual Event was terminated at 0036 on August 2, 2009. This call is being made due to the termination of the Unusual Event declared on July 30, 2009. An Event Summary Report is required by Exelon procedures within 24 hours of termination of the Unusual Event and will be communicated to the Headquarter Operations later today. The initial event was the result of the actuation of the SAT sudden pressure relay. When the transformer tripped, a slow automatic bus transfer resulted. When the RCPs (Reactor Coolant Pump) and condensate pumps were reenergized, they tripped on overcurrent causing the reactor trip. The sudden pressure relay has subsequently tripped during testing and may have caused the initial event. The licensee reported no damage to the plant. The licensee notified the NRC Resident Inspector. Notified the R3DO (Daley), IRD (McDermott), NRR (Howe), DHS (An), and FEMA (Biscoe).
The Event Summary Report was received and documented the following technical conclusions: The Unusual Event declaration was caused by a sudden pressure relay on SAT 242-1 causing a lockout of both SATs followed by a trip of Unit 2 due to the 2C RCP tripping during the automatic bus transfer for bus 258. This led to a loss of offsite power to Unit 2. It is currently unknown why the sudden pressure relay on SAT 242-1 actuated. Troubleshooting on the sudden pressure relay is in progress. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Daley). Notified the IRD (McDermott) and NRR (Howe) via e-mail. | |
| ENS 45605 | 4 January 2010 15:30:00 | 10 CFR 26.719, FFD Reporting requirements | Possesion of Alcoholic Beverage in Protected Area | At approximately 0930 on January 4, 2010 a gift package was brought into the protected area that contained a bottle of an alcoholic beverage. The individual that brought the package into the protected area was unaware that a bottle of an alcoholic beverage was in the gift package. The gift package was delivered to the intended individual. As the individual was reviewing the contents of the package, the bottle of alcoholic beverage was identified as part of the package. Security was immediately contacted and took possession of the bottle. The bottle was still sealed with the original seal. The gift package was always attended while inside the protected area. A fitness for duty evaluation of the personnel involved was conducted with no concerns identified. The individual was a non-licensed licensee supervisor. The licensee notified the NRC Resident Inspector. | |
| ENS 45618 | 10 January 2010 01:51:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Declared Due to a Fire in the Auxiliary Building Lasting Greater than 15 Minutes | At 1925 (hrs. CST), the Main Control Room received a notification of smoke in the Auxiliary Building Ventilation Supply Plenum. Fire Brigade and the Incident Commander were dispatched. Once (they) arrived at the scene, they noted smoke in the area of the 0C VA Supply Fan and requested the fan be shutdown. The 0C VA supply fan was shutdown at 1933 hrs. The operators reported smoke and a small fire coming from the inboard bearing of the 0C VA supply fan. A CO2 fire extinguisher was used to put the fire out and cool the bearing. A total of 2 extinguishers were used. The fire was declared out at 1941 hrs. Both Units remained stable and at full power during the entire event. There were no injuries, and no off site assistance was required. The licensee notified state and local authorities and the NRC Resident Inspector.
At 2111 CST (on) 1-9-2010, the Unusual Event was terminated. The fire is out. Both Units are stable. A fire watch is established. There are no signs of reflash. (Braidwood) no longer (meets the) Unusual Event action level threshold. The licensee has notified the NRC Resident Inspector. Notified IRD (Grant), R3DO (Stone), NRR EO (Giiter), DHS (Inzler) and FEMA (Casto). | |
| ENS 46021 | 17 June 2010 10:19:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Containment Spray Recirc Sump Isolation Valve Failure to Stroke Closed | At 0519 CDT on June 17th, Unit 2 was closing the 2CS009B, Containment Spray Recirc Sump Isolation Valve, as part of post maintenance testing when the valve stopped stroking (i.e. mid position). The 2CS009B valve was being stroked closed for restoration from a successful timed stroke in the open direction. The 2CS009B valve was manually closed and verified closed via limit switch indication. With the 2CS009B valve unable to be closed from the Main Control Room, an unanalyzed condition may have existed where, during a large break LOCA requiring cold leg recirc, the Refueling Water Storage Tank (RWST) had an additional flow path to the containment recirc sump. This potentially challenges the operators to complete the switchover prior to the RWST reaching 9%, the point at which pumps taking a suction from the RWST only are shutdown. This condition is still being evaluated. The licensee notified the NRC Resident Inspector. * * * RETRACTION FROM P. MOODY TO P. SNYDER AT 0404 ON 2/1/11 * * * At 0509 on June 17, 2010, Unit 2 was closing the 2CS009B, Containment Spray (CS) Recirculation Sump Isolation Valve, as part of post maintenance testing when the valve stopped stroking (i.e., mid-position). The 2CS009B was being stroked closed for restoration from a successful timed stroke in the open direction. The 2CS009B was manually closed and verified closed via limit switch indication. With the 2CS009B unable to close from the Main Control Room, an unanalyzed condition may have existed where, during a large break LOCA requiring cold leg recirculation, the Refueling Water Storage Tank (RWST) had an additional flow path to the recirculation sump. This potentially challenged the operators to complete the switchover prior to the RWST reaching 9%, the point at which pumps taking a suction from the RWST only are shutdown. While this condition was being evaluated, an ENS notification was made per ENS 46021 under 10CFR50.72(b)(3)(ii)(B). As the evaluation approached the 60-day reporting period, LER 2010-002 was issued in accordance with 10 CFR 50.73(a)(2)(ii)(B), assuming the results would yield an unanalyzed condition. Since then, an evaluation was completed. The results concluded the operators would have performed the switchover steps within the allowed time, before reaching the RWST empty alarm set point. Therefore, the Emergency Core Cooling System (ECCS) and CS system would have performed their design functions. The evaluation also determined the RWST outflows with 2CS009B in the open position during the ECCS switchover sequence did not affect the RWST vortex analysis. Based on no loss of design function, the plant was not in an unanalyzed condition and this event is not reportable per 10CFR 50.72(b)(3)(ii)(B) or 10CFR 50.73(a)(2)(ii)(B). This event was screened for additional reportability criteria contained in the Exelon Reportability Manual. Again, since there was no loss of design function there is no reportability requirement. Therefore ENS notification 46021 is being retracted. The licensee notified the NRC Resident Inspector. | |
| ENS 46178 | 16 August 2010 07:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trips at Both Units | Braidwood Unit 2 automatically tripped at 0206 (CST) due to a turbine generator trip due to generator lockout relay actuation. All systems responded as expected, with the auxiliary feed water pumps starting on Low-2 Steam Generator level. The Unit is stable in Mode 3, all primary systems are stable with the secondary heat sink being maintained via aux feed water and the steam dumps. Offsite power is supplying Unit 2, and both emergency diesel generators are available. Cause of generator lockout is under investigation. Braidwood Unit 1 automatically tripped at 0219 (CST) on a turbine trip caused by a loss of condenser vacuum. All systems responded as expected, with the auxiliary feed water pumps supplying steam generator levels. Secondary heat sink is steam generator PORVs. One steam generator safety valve is not fully seated. No steam generator tube leakage. Cause of the loss of vacuum is under investigation. For both Units all control rods fully inserted. There were no complications during the trip and all systems functioned as required. Offsite power remained available throughout the transient. The licensee notified the NRC resident inspector. Braidwood Unit 1's loss of condenser vacuum was caused by the loss of an electrical bus supplying the circ water pumps. At the time of this report, both plants were in a normal shutdown electrical lineup with the exception of the deenergized bus supplying power to the circ water pumps on Unit 1. The steam generator safety valve that has not fully seated was characterized as weeping a small amount of steam. The licensee is uncertain if the Unit 1 trip is related to the Unit 2 trip. | |
| ENS 46203 | 24 August 2010 16:40:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Essential Service Water Placed in a Line-Up That May Have Prevented Its Safety Function | At 1140 (CDT) on August 24th, Unit 2 received Essential Service Water (SX) discharge header pressure low and SX strainer delta pressure high alarms indicative of high flow. At the time, a 2B SX ASME surveillance was in progress which involved field operations by Equipment Operators (EO's). At the time of the event, SX discharge header pressure dropped to 65 psig, less than the 89 psig necessary for operability. The Control Room responded by directing the EO's to restore SX discharge header pressure, which was promptly restored. The 2B SX ASME surveillance sets initial conditions prior to data collection. The surveillance has the total SX flow be adjusted to 24000 gpm via the U2 Component Cooling Water (CC) heat exchanger outlet throttle valve, 2SX007. The subject flow was intended to be measured via an installed ultrasonic flow gauge 2FE-SX147. The EO's, instead used the U2 CC heat exchanger flow gauge 2FE-SX031. As a result, in an attempt to achieve 24000 gpm through the U2 CC heat exchanger, total SX flow exceeded the 24000 gpm since the U2 CC heat exchanger is but one of many loads the 2B SX pump is serving. For the 5 minutes described above, the SX system was in a lineup that may have prevented it to fulfill its safety function and placed Unit 2 in a potentially unanalyzed condition. This condition is still being evaluated. Site Engineering has determined no runout conditions existed. The licensee notified the NRC Resident Inspector.
The evaluation of the condition has been completed. Based on Essential Service Water (SX) system flow model runs performed, the conditions that existed at the time of the low SX header pressure would have resulted in low flow supplied to multiple safety related components. However, the safety function to provide necessary cooling to required safety-related safe shutdown equipment would have been met under design basis conditions with the auto-start of the 2A SX pump. Therefore, this did not result in a condition that could have prevented fulfillment of a safety function or in an unanalyzed condition that significantly degraded plant safety. Therefore, ENS notification 46203 is being retracted. The licensee notified the NRC Resident Inspector. Notified the R3DO (Daley). | |
| ENS 46262 | 20 September 2010 22:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 1704 CDT, Braidwood Unit 1 experienced an automatic reactor trip. The reactor trip red first out was Over Temperature Delta Temperature (OTDT). At the time of the reactor trip, the Instrument Maintenance Department was performing a calibration of Power Range Channel N-43 and a calibration of the 1C S/G Narrow Range Level Channel 1L-0538. The cause of the trip is unknown at this time. After the reactor trip occurred, all four Steam Generators reached their Low-2 reactor trip setpoint and Pressurizer pressure reached its low pressure reactor trip setpoint which is an expected response on a trip from full power. Steam Generator levels and Pressurizer pressure have been restored. Both the 1A and 1B Auxiliary Feedwater pumps auto started on the Low-2 Steam Generator levels as expected. All control rods fully inserted into the core. Train B Main Control Room Filtration system shifted to makeup mode and the Train B Fuel Handling Building ventilation shifted to Emergency Mode due to a spurious actuation signal. No secondary relief valves lifted and no secondary steam (was) released as a result of the reactor trip. The Main Steam Dumps are in service to the Main Condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. This report is being made per 10CFR50.72(b)(2)(iv)(B) for RPS actuation, 4-hr. notification, and per 10CFR50.72(b)(3)(iv)(A) for automatic actuation of the Auxiliary Feedwater system, 8-hr. notification. AC power is being provided by offsite power with the Diesel Generators in standby and all safety systems available. There is no Unit 2 impact. The licensee notified the NRC Resident Inspector. The licensee also anticipates that there will be a press release issued regarding this event. | |
| ENS 46414 | 12 November 2010 14:18:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Outside Air Intake Noble Gas Channel Setpoints Non-Conservative | At 0818 CST, 11/12/10, Radiation Protection determined that the setpoints for the Control Room Outside Air Intake Noble Gas channels are non-conservative. This affects Tech Spec 3.3.7 required monitors 0PR31B, 0PR32B, 0PR33B and 0PR34B (Noble gas channels). The current setpoints for 0PR31B, 0PR32B, 0PR33B and 0PR34B are High Alarm 9.55E-05 microCi/cc and Alert Alarm 9.55E-06 microCi/ml. The calculated required setpoints are High Alarm 6.61E-05 microCi/cc and Alert Alarm 6.61E-06 microCi/cc. This is approximately a 30% decrease in setpoints from the current setpoints. LCO 3.3.7 conditions A and B were entered at 0818, 11/12/10. LCO 3.3.7 conditions A and B required actions completion time is 1 hour. All required actions were complete at 0900 11/12/10, less than 1 hour. This report is being made per 10 CFR 50.72(b)(3)(v), Event or condition that could have prevented fulfillment of a safety function, 8 hour non-emergency notification. Both trains of radiation monitor setpoints were non-conservative and actuate control room ventilation in emergency mode. The margin available in the control room dose analysis will be reviewed to confirm impact on safety function. The setpoints have been non-conservative since 1999. The licensee notified the NRC Resident Inspector.
At 0818 CST, 11/12/2010, Radiation Protection determined that the setpoints for the Control Room Outside Air Intake Noble Gas channels are non-conservative. This affects Tech Spec 3.3.7 required monitors 0PR31B, 0PR32B, 0PR33B, 0PR34B (Noble Gas channels). ENS notification was made under ENS 46414 under 10 CFR 50.72(b)(3)(v)(D). The design basis accidents do not credit automatic actuation of the Control Room Outside Air Intake system to the Emergency mode from a high radiation signal. Therefore, the high radiation signal is not needed to mitigate the consequences of an accident and this event did not result in a safety system functional failure. Therefore, ENS notification 46414 is being retracted. The licensee notified the NRC resident inspector. Notified R3DO (Bloomer) | |
| ENS 46415 | 12 November 2010 19:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Inaccurate Information Provided in License Amendment Request | At 1300, on November 12, 2010, Exelon Generation Company LLC concluded that inaccurate information contained in the PRA technical bases for a 1987 License Amendment Request (LAR) for Byron and Braidwood Stations would have potentially impacted the acceptability of the LAR by the NRC. The LAR was to extend Allowed Outage Times (AOT) from 72 hours to 7 days for several systems, to include the Component Cooling (CC) and Residual Heat Removal (RH) Systems. The original design intent of the CC system was that each unit has two independent CC pumps and a fifth pump (U0) CC pump could be used as an operable spare for any of the unit specific pumps. This is how CC was modeled in the PRA technical justification for the 1987 LAR. However, a piping configuration design flaw that was recently evaluated in that the U0 CC pump could not be considered an operable spare for either unit's B pumps was not correctly modeled in the PRA. During the evaluation to assess the potential significance of this CC design flaw on the PRA justification for the 1987 LAR, another potentially significant discrepancy was discovered in that it appears the operational practice to always split CC trains after a design basis LOCA was not modeled correctly in the RH analysis. Administrative controls have been put in place to restrict the AOT for the CC pumps and RH trains to the pre-LAR timeframe of 72 hours pending the permanent corrective actions. In addition, administrative controls have been put in place to prohibit the U0 CC pump from being an operable spare for either unit's B trains. This event is being reported as an unanalyzed condition that significantly degrades plant safety under 10 CFR 50.72(b)(3)(ii). The NRC Resident Inspectors have been notified | |
| ENS 46549 | 18 January 2011 12:00:00 | 10 CFR 26.719, FFD Reporting requirements | Confirmed Positive Test for Illegal Drug | A non-licensed employee supervisor has a confirmed positive test for an illegal drug during a random fitness for duty test. The employees access to the plant has been cancelled. Contact the Headquarters Operations Officer for more details. The NRC Resident Inspector has been notified. | |
| ENS 46694 | 24 March 2011 15:18:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unexpected Loss of Annuciators During Planned Maintenance | During a planned maintenance activity on the Unit 2 main control room alarm cabinets, it was identified that all Unit 2 safety system annunciators were lost. This was identified at 1006 (CDT). Main Control Board indicators remained functional. At 1018 (CDT), the Shift Manager declared an Unusual Event under Emergency Action Level MU6. This was due to an unplanned loss of most (approximately 75%) safety system annunciators for > 15 minutes. The planned maintenance activity was not expected to affect the amount of annunciators that were lost. At 1030 (CST), the (planned maintenance) clearance order was cleared and power was restored to the Unit 2 annunciators. There was not transient on other plant equipment and the plant remained stable before and after this event. The cause for the unexpected loss of annunciators is not clearly understood and is still under investigation. The licensee notified the NRC Resident Inspector.
The Unusual Event was terminated at 1047 CDT on 03/24/11. All annunciators have been restored and an investigation will be conducted to determine the cause. Notified R3DO (Cameron) | |
| ENS 46707 | 30 March 2011 01:00:00 | 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 | Potential Voiding in Auxiliary Feedwater Alternate Suction Line | The design of the Auxiliary Feedwater (AF) system is for the AF pumps to normally take suction from the condensate storage tank. If the condensate storage tank is not available, the essential service water system provides the alternate supply. Due to the AF system suction piping and valve configuration, a voided section of pipe could exist in the portion that isolates the condensate storage tank supply from the essential service water supply. A preliminary vendor analysis has determined that the void fraction to reach the pump in a dynamic scenario exceeds the acceptance criteria for AF pump operability. Based on past operation in this configuration, the event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). Further review of the void model and pump performance characteristics are planned. In 2011, prior to the completion of this analysis, the void was refilled and verified full for the 'B' trains at Braidwood U1 and U2. Filling the voided piping of both 'A' trains at Braidwood U1 and U2 is in progress. Once filled, the AF systems are operable. The licensee has notified the NRC Resident Inspector. | |
| ENS 46712 | 30 March 2011 20:38:00 | Other Unspec Reqmnt | Discovery of After-The-Fact Emergency Condition (Unusual Event) | An extent of condition review of Braidwood Unit 2 unplanned loss of safety system annunciators Emergency Plan Unusual Event on March 24, 2011 (ENS number 46694) was performed for both Units of Braidwood Station. During this review it was identified that a previous unknown loss of annunciators had also occurred on August 10, 2010 from 1024 to 1136 CT on Unit 2. This condition occurred during planned maintenance on annunciator cabinet 2PA19J power supply capacitors. The maintenance performed on August 10, 2010 would normally not cause a loss of all Unit 2 annunciators. During the work, it was expected to lose approximately one third of the annunciators. Latent annunciator system problems identified from the March 24, 2011 event caused a loss of all Unit 2 annunciators and contributed to this condition being unknown to Main Control Room operators. All Unit 2 indications and computer points to the sequence of events recorder remained available and Unit 2 was stable during this timeframe. At 1538 CT on 3/30/11, it was determined that the August 10, 2010 condition met the threshold for Emergency Action Level MU6, UNPLANNED loss of most or all safety system annunciation or indication in the control room for greater than 15 minutes. This notification is being made as an undeclared Unusual Event Emergency Plan Classification per 10 CFR 50.72(a)(1)(ii). Per NUREG 1022, a 1- hour notification is required when a condition existed which met the emergency plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of the discovery. The licensee notified the NRC Resident Inspector. | |
| ENS 46766 | 19 April 2011 04:09:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Plant Process Computer Removed from Service for Planned Replacement | At 2309 (CDT) on April 18, 2011, the Unit 2 Plant Process Computer (PPC) was removed from service for planned replacement in the current Unit 2 refueling outage. The Unit 2 PPC feeds the Safety Parameter Display System (SPDS) used in the Main Control Room (MCR) and the Technical Support Center (TSC). The Unit 2 PPC also feeds the Emergency Response Data System (ERDS). The Unit 1 and Unit 2 PPCs also feed the Plant Parameter Display System (PPDS) used in the MCR, TSC and Emergency Operations Facility (EOF). Meteorological data will remain available. The dose assessment program will remain functional as the Unit 1 Plant Process computer will be capable of providing the necessary data through PPDS to run the program. The dose assessment program is not affected by the Unit 2 PPC being out of service. As compensatory measures, the backup method to fax or communicate via a phone circuit applicable data to the NRC, TSC and EOF exists. There is no impact to the Emergency Notification System (ENS) or Health Physics Network (HPN) communications systems. The new Unit 2 PPC is scheduled to be functional on April 23, 2011. However, based on the mode Unit 2 will be in, this will limit the number of points that would provide usable data. The Unit 2 PPC will be tested as mode changes occur. The Unit 2 PPC is planned to be declared functional by Mode 2. A follow-up ENS call will be made once the Unit 2 PPC is declared functional. The loss of SPDS and ERDS is a 'major loss of assessment capability' and is reportable under 10CFR50.72(b)(3)(xiii). The NRC Senior Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS call.
As of 0815 CT on May 10, 2011, the Unit PPC is considered operational with respect to the Safety Parameter Display System (SPDS), Plant Parameter Display System (PPDS) and Emergency Response Data System (ERDS). Therefore, a Major Loss of Assessment Capability no longer exists on Unit 2. The EP Manager will contact the NRC Computer Center today to conduct an ERDS test for Unit 2 to ensure the data is being satisfactorily sent to the NRC. The NRC Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS update. Notified R3DO(Passehl) | |
| ENS 46868 | 20 May 2011 23:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Marine Life Inside Afw Piping | The past operability of the 2A train of Auxiliary Feedwater has been called into question based on finding clam shells in the safety related water source piping. During performance of valve strokes on 5/9/2011 per an Operations Department surveillance, clam shells were found while draining a section of the safety related water source piping. System Engineering collected the shells as part of troubleshooting. Based on analysis performed by System Engineering, the 2A Train of Auxiliary Feedwater was not operable with clam shells in the pipe. The amount of shells present would have caused an unacceptable differential pressure across the 2A train Auxiliary Feedwater System flow control valves. The extent of condition has been evaluated for the other Auxiliary Feedwater trains for both units and it has been determined that the only affected train is 2A. The 2B Auxiliary Feedwater train has also been inoperable at various times over the past 3 years for maintenance. The clams were flushed out of the 2A Train Auxiliary Feedwater suction piping during a recent refueling outage and both trains of Auxiliary Feedwater on Unit 2 are operable. The licensee has not implemented any compensatory measures nor are they in any LCO's as a result this event. The NRC Resident Inspector has been notified. | |
| ENS 46936 | 8 June 2011 15:26:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Due to Flooding That Potentially Affected Safety Related Equipment | Flushing activities were in progress on the 2A Auxiliary Feedwater (AF) Pump suction line from the Essential Service Water System (SX). At 1011 (CDT), during this flushing activity, the flushing hose ruptured and caused flooding in the Auxiliary Building that had the potential to affect safety related equipment needed for the current operating mode. This was due to the flood waters contacting the motors of the 1A and 2A AF Pumps. At 1026, an Unusual Event was declared by the Shift Emergency Director because the conditions for EAL entry were met for EAL HU5. Specifically the EAL conditions were 'Flooding in the Auxiliary Building that has the potential to affect safety related equipment needed for the current operating mode.' Auxiliary Feedwater is required to be operable in Mode 1 for each unit. The leak was immediately isolated (within 45 seconds of hose rupture) and remains isolated. Maintenance personnel are in the process of testing the 1A and 2A AF Pump motor to determine operability. The appropriate Tech Spec Action Statements have been entered on each Unit for the AF Pumps. The leak also caused the wetting of MCC (Motor Control Center) 132X1, which feeds the 1B AF Pump control power. The Unit 0 (Common) Component Cooling Pump feed breaker cubicle was also wetted. Investigation into these components are also in progress. The licensee has entered Technical Specification LCO 3.7.5 Condition A which requires restoration of the one of the AF pumps within 72 hrs. There were no personnel injuries. The licensee has notified the NRC Resident Inspector.
At 1122 (CDT) on 6/8/11, the Unusual Event was terminated due to the flooding conditions no longer existing. Evaluation of wetted components are still in progress. The NRC Resident Inspector has been notified of event. Notified IRD (Gott), R3DO (Lara), NRR EO (Thorp), FEMA (Blankenship), and DHS (Flinter). | |
| ENS 46942 | 9 June 2011 19:22:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Inadvertent Siren Activation Due to Lightning Strike | This siren affects both Dresden and Braidwood Stations. At approximately 1422 (CDT) on June 9, 2011, a Grundy County representative notified Exelon of an inadvertent siren activation. The Exelon representative contacted the siren contractor at 1423 (CDT) inquiring about siren activity in the shared Braidwood/Dresden Emergency Planning Zones. Upon review and polling of the system at 1444 (CDT), siren BD11 was identified in the area of concern with a communication failure. The siren contractor was immediately dispatched and arrived at the siren location at approximately 1520 (CDT). Upon arrival, the siren was not sounding and significant damage was identified, which was indicative of a direct lightning strike. The siren was declared out of service and repairs will begin on June 10th. The licensee notified the NRC Resident Inspector and Grundy County local authorities. See Dresden Event #46943 for same event notification. | |
| ENS 47319 | 4 October 2011 21:35:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Onsite Fatality | At 1430 CDT, 10/4/11, the licensee Main Control room received an emergency phone call for medical assistance. The caller reported that he had a person down in the auxiliary building. On-site medical response team reported the person was not breathing and not responsive. The symptoms were non-occupational. The person was transported off site via ambulance to the local hospital. At 1635, 10/4/11, the licensee was informed that the individual was declared deceased. OSHA is being notified pursuant to the requirements of 29CFR1904.39. This ENS report is being made in accordance with 50.72 (b)(2)(xi) There was no radioactive contamination involved in this event. The licensee does not plan any media or press release and have not notified any other government agencies besides OSHA. The NRC Resident Inspector will be notified. | |
| ENS 47622 | 26 January 2012 03:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Out of Service Due to Planned Maintenance | Planned preventive maintenance activities are being performed on the Braidwood Nuclear Station Technical Support Center (TSC) Ventilation System. These work activities are planned to be performed and completed expeditiously within 8 hours. This maintenance activity includes the performance of preventive maintenance on the TSC outside air supply fan unit which affects the TSC emergency filter train and air handling unit. During a portion of the time these activities are being performed, this equipment will not be available for operation. As such, the TSC Ventilation will be rendered non-functional during the performance of portions of the work activity. If an emergency condition occurs that requires activation of the Technical Support Center, during the time this work activity is being performed, it will take no more than 4 hours to return the equipment back to functional status, dependent on the stage of the work activity at the time an emergency occurs. Plans are to utilize the TSC for any declared emergency during the time this work activity is being performed as long as radiological conditions allow. This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2 since this work activity affects an emergency response facility for the duration of the maintenance. The licensee notified the NRC Resident Inspector.
Braidwood Nuclear Station TSC ventilation was restored to available status at 0635 CST on January 26, 2012. The previously reported system preventative maintenance has been completed. The licensee notified the NRC Resident Inspector. Notified R3DO (L. Kozak). | |
| ENS 47637 | 3 February 2012 06:00:00 | Other Unspec Reqmnt | VOLUNTARY REPORT - DESIGN VULNERABILITY IN 4.16kV BUS UNDER-VOLTAGE SCHEME On January 30, 2012, a design vulnerability was discovered at Byron and Braidwood stations in the Engineered Safety Feature 4.16kV bus under-voltage protection scheme for Braidwood Station Units 1 and 2. Specifically a voltage unbalance created by an open circuit of either the A or C phase from the offsite grid to the System Auxiliary Transformers (SAT) is not designed to actuate the protective relays on the 4.16kV safety bus that provides isolation from the offsite grid and the automatic start and loading of the emergency onsite diesel generators. Two under-voltage relays are provided on each 4.16kV safety bus, which are combined in a two out of two logic to generate a loss of power signal. The relays are sensing voltage between two phases (i.e., A&B and B&C). An open circuit condition on the C phase or the A phase would not satisfy the two out of two logic. This condition results in both 4.16kV safety buses remaining energized with a bus undervoltage situation and results in equipment protective devices actuating from over-current conditions. This configuration is a non-conforming condition in that the design of the under-voltage relays and logic was intended to identify degraded grid conditions, not loss of a single phase. With an open circuit on the A or C phase from the grid to the SATs, during normal operations, operators have to diagnose the condition and manually isolate safety buses from offsite power which would automatically start and load the emergency diesel generators. During a design basis event concurrent with an open circuit on A or C phase from the grid to the SATs, analysis performed to date indicates that starting of the ECCS loads would have caused the bus voltage to decrease sufficiently to actuate the under-voltage protective relays and restore cooling with emergency onsite power without challenging fuel design limits. The 4.16kV safety bus under-voltage protection scheme at Byron and Braidwood is believed to be a typical industry design. This design issue is being evaluated at the other Exelon stations. The results of this evaluation will be shared with the NRC. Therefore, this condition is being reported as a voluntary notification due to its potential generic industry applicability." The licensee notified the NRC Resident Inspector. | ||
| ENS 47845 | 18 April 2012 02:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Plant Process Computer Removed from Service for Planned Replacement | At 2130 (CDT) on April 17, 2012, the Unit 1 Plant Process Computer (PPC) was removed from service for a planned replacement in the current Unit 1 Refueling Outage. The Unit 1 PPC feeds the Safety Parameter Display System (SPDS) used in the Main Control Room (MCR) and the Technical Support Center (TSC). The Unit 1 PPC also feeds the Emergency Response Data System (ERDS). The Unit 1 and Unit 2 PPCs also feed the Plant Parameter Display System (PPDS) used in the MCR, TSC and Emergency Operations Facility (EOF). Meteorological data will remain available in the MCR but not through ERDS for either Unit 1 or Unit 2. The dose assessment program will remain functional as the Unit 2 Plant Process computer will be capable of providing the necessary data through PPDS to run the program. The dose assessment program is not affected by the Unit 1 PPC being out of service. As compensatory measures, a proceduralized backup method to fax or communicate via a phone circuit applicable data to the NRC, TSC, and EOF exists. There is no impact to the Emergency Notification System (ENS) or Health Physics Network (HPN) communication systems. The new Unit 1 PPC is scheduled to be functional on April 21, 2012. However, based on the mode Unit 1 will be in, this will limit the number of points that would provide usable data. The Unit 1 PPC will be tested as mode changes occur. The Unit 1 PPC is planned to be declared functional by Mode 2. A follow-up ENS call will be made once the Unit 1 PPC is declared functional. The loss of SPDS and ERDS is a 'major loss of assessment capability' and is reportable under 10CFR50.72(b)(3)(xiii). The NRC Senior Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS call.
As of 1035 EDT on May 15, 2012, the Unit 1 PPC is considered operational with respect to the Safety Parameter Display System (SPDS), Plant Parameter Display System (PPDS) and Emergency Response Data System (ERDS). Therefore, a Major Loss of Assessment Capability no longer exists on Unit 1. The EP Manager will contact the NRC Computer Center for Unit 1 to ensure the ERDS data is being satisfactorily sent to the NRC. The NRC Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS update. The R3DO (Patty Pelke) has been notified. | |
| ENS 47934 | 23 April 2012 06:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Ultrasonic Examination Results in an Indication on the Reactor Pressure Vessel Head Penetration | On April 23, 2012, during the Braidwood Station Unit 1 refueling outage, it was determined that the results of planned ultrasonic (UT) examinations performed on one penetration (Penetration No. 69) did not meet the applicable acceptance criteria, requiring repair prior to returning the reactor pressure vessel head to service. A portion of the indication was conservatively assumed to be within the J-groove weld. At the time of the examination, the Braidwood Station Unit 1 reactor pressure vessel head was classified as a low susceptibility head. The cause of the recordable indication is attributed to Primary Water Stress Corrosion Cracking. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to ensure the structural integrity of the reactor pressure vessel head pressure boundary. All of the remaining penetrations have been examined during the current refueling outage. Repairs to Penetration No. 69 were completed prior to commencing startup. No other repairs were required. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. The NRC Resident Inspector has been notified. As a result of a follow-up review, at 0600 CDT on 05/18/12, the licensee determined this was an 8 hour reportable event while Unit 1 was in Mode 4. The licensee did not determine this was a reportable event immediately after the UT was conducted on April 23, 2012. The unit is currently in Mode 3. Reactor Vessel Head Penetration No. 69 is a control rod penetration. Reactor Coolant Leakage test are satisfactory and there are no indications of boron buildup on the Reactor Vessel Head. Control rod testing has not yet been performed. |