Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 52802 | 12 June 2017 14:14:00 | Other Unspec Reqmnt 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification for Discharge of Circulating Water | On Sunday, June 11, 2017 at 1200 CDT, a Chemistry Technician reported water from the Circulating Water Blowdown (CW B/D) valve building had been pumped onto the ground outside the CW B/D building during a maintenance activity. The sump in the CW B/D building has a sump pump discharge hose that had been routed outside to the ground near the outfall canal rather than to the permitted outfall path. Water sample analysis results confirmed the presence of tritium. The sump pump discharge hose was rerouted to the required outfall on June 11th, 2017 at 1500 CDT. The discharge flow did not leave the site boundary before reaching the permitted discharge pathway. In accordance with the Braidwood Station Illinois Environmental Protection Agency (IEPA) Consent Order dated March 11, 2010, the IEPA was notified of the blowdown line release on June 12, 2017 at 0914 CDT; the Illinois Emergency Management Agency (IEMA) was subsequently notified. Local agencies in the Braidwood area will also be notified. Follow-up analyses are being performed to determine if additional IEPA or IEMA notifications will be required. The release tank values for other radionuclides (i.e., other than tritium) were interpolated based on tritium values for the release tank, dilution from the CW B/D, and sampled sump tritium values; and determined to be less than Minimum Detectable Activity (MDA). Within 4 hours of determination that state and local agencies notification will be made, the NRC Operations Center is required to be notified. There were no public health risks associated with the event. The NRC Resident Inspector was notified. Due to the notification of government agencies, this event is being reported under 10 CFR 50.72(b)(2)(xi).
The Illinois Environmental Protection Agency (IEPA) and the Illinois Emergency Management Agency (IEMA) were notified on Wednesday, June 14, 2017, at 1206 CDT and 1213 CDT, respectively, of the preliminary results of estimated quantity of curies of tritium release (approximately 0.009 curies of tritium) and the estimated volume of the release (approximately 35,000 gallons). The discharge flow did not leave the site boundary before reaching the permitted discharge pathway. The NRC Resident Inspector was notified of the update to IEPA and IEMA. Notified R3DO (Stone).
The Illinois Environmental Protection Agency (IEPA) and the Illinois Emergency Management Agency (IEMA) were provided with a follow-up report on June 16, 2017, at 1528 CDT as required by 35 IAC 1010.204. The report includes the estimated quantity of curies of tritium released (approximately 0.009 curies), the estimated volume of the release (approximately 35,000 gallons), and actions taken in response to the on-site release (installation of groundwater monitoring wells, implementation of a sampling program to monitor the concentrations and migration, if any, of the tritium in the groundwater on-site, and the installation of a remediation system). The NRC Resident Inspectors have been notified of the update to IEPA and IEMA. Notified R3DO (Stone).
On 6/20/2017, the Illinois Environmental Protection Agency (IEPA) Des Plaines Regional Office was sent a follow-up report pursuant to the Consent Order, dated March 11, 2010. This report contains that same information as the follow-up report required by 35 IAC 1010.204 that was submitted to the IEPA and the Illinois Emergency Management Agency on June 16. 2017. The NRC Resident Inspector has been notified of the update to IEPA. Notified R3DO (Pelke). | ||
| 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. | Emergency Diesel Generator | ||
| 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 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. | Feedwater | |
| 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. | Feedwater | |
| 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 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) | Feedwater |