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| Entered date | Site | Region | Reactor type | System | Scram | Event description | |
|---|---|---|---|---|---|---|---|
| ENS 56875 | 29 November 2023 17:02:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: This notification is being made per 10 CFR 50.72(b)(2)(xi), as a result of notifications made to State and local government agencies for the discovery of an oil sheen in the discharge canal outside Unit 3. The New York State Department of Environmental Conservation and Westchester County Department of Health were notified. No sheen was observed in the river or at the southern end of the discharge canal near the outfall gates. Clean up efforts are underway. The licensee will notify the NRC Project Manager. | ||
| ENS 56866 | 20 November 2023 17:53:00 | Dresden | NRC Region 3 | GE-3 | High Pressure Coolant Injection Emergency Core Cooling System | The following information was provided by the licensee via email: At 0956 (CST) on November 20, 2023, accumulated gas was identified in the Dresden Unit 2 high pressure coolant injection (HPCI) system discharge header. As a result, the HPCI system was declared inoperable. Since HPCI is a single-train system, this is a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The HPCI system was subsequently vented, and the accumulated gas has been removed, restoring the Dresden Unit 2 HPCI system to an operable status. All other emergency core cooling systems remained operable during this time period. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee administratively verified the isolation condenser was operable after declaring HPCI inoperable as required by technical specifications. The licensee stated there was no increase in plant risk. The cause of gas accumulating in the Dresden Unit 2 HPCI discharge header is under investigation, and this issue has been entered into the licensee's corrective action program. | |
| ENS 56865 | 19 November 2023 14:44:00 | Quad Cities | NRC Region 3 | GE-3 | The following information was provided by the licensee via phone and fax: On November 18, 2023, the presence of alcohol was discovered inside the protected area. In accordance with the Constellation Fitness For Duty (FFD) Program, the individual has been escorted offsite and access to the plant denied pending the results of an investigation. This event is being reported under 10 CFR 26.719(b)(1) as it represents a significant FFD violation. The NRC Resident Inspector has been notified. | ||
| ENS 56864 | 19 November 2023 00:53:00 | Harris | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 2138 EST on November 18, 2023, Harris Nuclear Plant notified the National Response Center of a biodegradable oil leak that entered the Harris Lake. The North Carolina Department of Environmental Quality will also be notified of this condition on November 19, 2023. The oil leak was less than one gallon and came from a temporary pump. The leak has stopped, and spill cleanup is underway. This condition did not violate any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a notification to another government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | ||
| ENS 56863 | 18 November 2023 02:51:00 | River Bend | NRC Region 4 | GE-6 | Feedwater Reactor Protection System Emergency Diesel Generator Control Rod | Manual Scram | The following information was provided by the licensee via phone and email: On November 17, 2023, at 2215 CST, River Bend Station (RBS) was operating at 30 percent reactor power performing plant startup activities when an isolation of low-pressure feedwater string `A' occurred. The team entered applicable alternate operating procedures and inserted control rods to exit the restricted region of the power to flow map. Feedwater temperature continued to lower until it challenged the prohibited region of the AOP-0007 graph requiring a reactor scram. The team inserted a manual reactor scram at 2355 from 24 percent reactor power. All control rods fully inserted and there were no complications. All systems responded as designed. Currently RBS Unit 1 is stable with reactor level being maintained 10 to 51 inches with feed and condensate, and pressure being maintained 500 to 1090 psig using steam drains. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) Specified System Actuation as result of Group 3 isolations. The NRC Senior Resident inspector has been notified. No radiological releases have occurred due to this event from the unit. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The electric plant is in a normal lineup for current plant conditions with all emergency diesel generators available. The cause of the initial isolation of low-pressure feedwater string "A" is still under investigation. |
| ENS 56861 | 16 November 2023 21:30:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | Emergency Diesel Generator HVAC Control Room Envelope Containment Spray | The following information was provided by the licensee via phone and email: 11/05/23, 2200 CST: Essential Chiller 'B' train and associated cascading equipment were declared INOPERABLE for planned maintenance. Unit 2 entered the Configuration Risk Management Program as required by Technical Specifications on 11/12/23 at 2200. 11/16/23, 1541: Essential Chiller 'C' train and associated cascading equipment were declared INOPERABLE due to an unexpected material condition causing the Essential Chiller to trip. The most limiting (Limiting Condition of Operability) LCO is 3.7.7, Action c. This condition resulted in the INOPERABILITY of two of the three safety trains required for the accident mitigating function including: High Head Safety Injection, Low Head Safety Injection, Containment Spray, Electrical Auxiliary Building HVAC, Control Room Envelope HVAC, Essential Chilled Water. This is an 8 hour reportable condition per 10CFR50.72(b)(3)(v)(D) because it could affect the ability to mitigate the consequences of an accident. A risk analysis was performed for the equipment INOPERABILITY and mitigating actions have been taken per site procedures. All 'A' train equipment remains operable. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The 'B' train Emergency Diesel Generator was also inoperable due to planned maintenance and continues to be inoperable. It was considered in the Configuration Risk Management Program and it was determined this condition could be maintained. LCO 3.7.7, Action c requires reactor shutdown within 72 hours. | |
| ENS 56858 | 16 November 2023 12:12:00 | Brunswick | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: At 0906 Eastern Standard Time (EST) on November 16, 2023, it was determined that a non-licensed employee supervisor failed a test specified by the Fitness for Duty (FFD) testing program. The individual's authorization for site access has been removed. The NRC Resident Inspector has been notified. | ||
| ENS 56856 | 16 November 2023 05:15:00 | Calvert Cliffs | NRC Region 1 | CE | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | Automatic Scram | The following information was provided by the licensee via email: At 0227 EST on 11/16/23, Calvert Cliffs Unit 2 experienced an automatic trip from the reactor protection system (RPS) based on reactor trip bus undervoltage (UV). At that time, a loss of U-4000-22 (13 kV to 4 kV transformer) caused a loss of 22, 23, and 24 4 kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV. The loss of 22 and 23 4 kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4 kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4 hour report ESFAS (engineering safety features actuation system) actuation (2B DG start on UV) is reportable under 10 CFR 50.72(b)(3)(iv)(A) - 8 hour report AFW operation is reportable under 10 CFR 50.73(a)(2)(iv)(A) - 60 day report The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. There was no impact on Unit 1 operations. Unit 2 is stable in mode 3.
ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8 hour report Notified R1DO (Defrancisco). |
| ENS 56852 | 14 November 2023 14:36:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | Manual Scram | The following information was provided by the licensee via phone and email: At 1041 CST on 11/14/23 with Farley Unit 2 in Mode 1 at 10 percent power, the reactor was manually tripped due to rising steam generator levels. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Auxiliary feedwater (AFW) was manually initiated in accordance with plant procedures and is feeding the steam generators. Heat removal is being provided via the atmospheric relief valves. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. The licensee attempted to take manual control of the feedwater control valves to lower steam generator level but, due to reaching a steam generator level that requires a manual trip, the licensee manually tripped the reactor. |
| ENS 56850 | 12 November 2023 22:02:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On November 12, 2023, at 0300 EST, a Watts Bar contractor was transported offsite for medical treatment due to a work-related injury. Upon arrival at an offsite medical facility, medical personnel determined the injury required the individual to be admitted into the hospital and will be kept overnight. The individual was inside of the Radiological Controlled Area, however was free released with no contamination. The injury and hospitalization were reported to the Occupational Safety and Health Administration (OSHA) under 29 CFR 1904.39(a)(2). The contracting agency informed OSHA at 1319 EST. Watt Bar Operations personnel were officially notified by the contracting agency of the report made to OSHA at 1945 EST. This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. | ||
| ENS 56849 | 11 November 2023 00:23:00 | Waterford | NRC Region 4 | CE | The following information was provided by the licensee via email: At 1545 CST on November 10, 2023, personnel at Waterford Steam Electric Station Unit 3 determined that 19 conduits in the engineered safety features actuation system (ESFAS) auxiliary relay cabinets A and B did not have the required fire seals for bay separation. This condition meets the criteria involving an unanalyzed condition that significantly affects plant safety. The plant is currently defueled. Decay heat is being removed by normal spent fuel cooling system operations. ESFAS is not required to be operable in the current plant mode. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. There was no impact to the health and safety of the public or plant personnel. The NRC Region 4 Branch Chief (Dixon) has been notified. | ||
| ENS 56848 | 10 November 2023 22:52:00 | South Texas | NRC Region 4 | Westinghouse PWR 4-Loop | HVAC Control Room Envelope Containment Spray | The following information was provided by the licensee via email: On 11/10/23 at 0642 CST, essential chiller 'B' train and cascading equipment was declared inoperable due to chill water temperature exceeding limits. At 1413 CST, essential chiller 'C' train and cascading equipment was declared inoperable due to discharge pressure exceeding limits. This condition resulted in an inoperable condition on two out of the three safety trains for the accident mitigating function including the 'B' and 'C' train high head safety injection, low head safety injection, containment spray, electrical auxiliary building HVAC, control room envelope HVAC, and essential chill water. All 'A' train equipment remained operable. This was determined to be reportable within 8 hours as required by 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Plant is in a 72 hour limiting condition for operation per technical specification 3.7.7. Restoration of 'B' train anticipated on 11/11/23 mid day. | |
| ENS 56846 | 10 November 2023 03:14:00 | Susquehanna | NRC Region 1 | GE-4 | Feedwater Reactor Protection System Main Condenser | Manual Scram | The following information was provided by the licensee via email: At 0118 EST, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually scrammed due to degrading main condenser vacuum. The scram was not complex, with all systems responding normally post-scram. The main turbine bypass valves opened automatically to maintain reactor pressure. Operations responded and stabilized the plant. Reactor water level is being maintained via feedwater pumps. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not impacted. Due to Reactor Protection System actuation while critical, this event is being reported as a four-hour and eight-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Unit 1 reactor is currently stable in mode 3. An investigation is in progress into the cause of the degrading condenser vacuum. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56845 | 9 November 2023 15:55:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a summary of information provided by the licensee via email: A controlled substance was found in the protected area. The NRC Resident Inspector has been notified. | ||
| ENS 56841 | 8 November 2023 13:27:00 | Calvert Cliffs | NRC Region 1 | CE | Steam Generator Feedwater Auxiliary Feedwater | The following information was provided by the licensee via phone and email: At 0645 EST, on November 8, 2023, with Unit 2 in Mode 3 at zero percent power, a manual actuation of the auxiliary feedwater system (AFW) occurred during a planned plant cooldown. The reason for the AFW manual-start was a trip of the 22 steam generator feed pump due to a high casing level. The 23 AFW motor driven pump was manually started in accordance with implementation of AOP-3G, Malfunction of Main Feedwater System to restore steam generator levels. There was no impact to Unit 1. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No other systems were affected. No other compensatory or mitigation strategies implemented. Plant cooldown was the only significant evolution in progress. No impact to other technical specifications or limiting conditions for operation. All systems functioned as required. The electric plant is being supplied by offsite power with all diesel generators available. No significant increase in plant risk. There was nothing unusual or not understood. | |
| ENS 56839 | 7 November 2023 18:42:00 | Calvert Cliffs | NRC Region 1 | CE | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | Automatic Scram | The following information was provided by the licensee via email: At 1617 on 11/7/2023, Calvert Cliffs Unit 2 experienced an automatic trip from a Reactor Protection System (RPS) based on reactor trip bus under voltage (UV). At that time a loss of U-4000-22 caused a loss of 22, 23, and 24 4kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV condition. The loss of 22 and 23 4kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4-hour report. ESFAS actuation (2B DG start on UV) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. Site Senior NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 was unaffected. Estimation of duration of shutdown is 24 hours. |
| ENS 56838 | 7 November 2023 18:18:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | Steam Generator Feedwater Auxiliary Feedwater | The following information was provided by the licensee via email: On November 07, 2023 at 1200 EST, it was discovered that all pumps in the Auxiliary Feedwater system were inoperable due to the loss of control power to the 'B' train Emergency Feedwater (EFW) flow control valve which supplies the 'D' steam generator. The redundant 'A' train EFW control valve for the 'D' steam generator remains functional, as well as the capability of the Auxiliary Feedwater system to supply all steam generators. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(B). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The 'A' and 'B' EFW Flow Control Valves are arranged in a series configuration for each Steam Generator. Failure of any of the 8 EFW Flow Control Valves to meet its Surveillance Requirements will render all EFW Pumps inoperable per tech specs. | |
| ENS 56834 | 5 November 2023 17:52:00 | Waterford | NRC Region 4 | CE | Steam Generator | The following information was provided by the licensee via email: At 1033 CST on November 5, 2023, while in a refueling outage, it was determined that Waterford Steam Electric Station, Unit 3, did not meet the performance criteria for steam generator structural integrity in accordance with Technical Specification 6.5.9.b.1, Steam Generator Program, due to two tube failures in the number 1 steam generator. The condition was identified during performance of in-situ pressure testing. The affected tubes will be plugged. The plant is currently stable with all fuel in the spent fuel pool. Decay heat is being removed by normal spent fuel cooling system operation. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A) as a degraded condition. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56832 | 4 November 2023 01:11:00 | Comanche Peak | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On 11/03/2023 at 2231 CDT, a security officer found 2 bottles of vanilla extract in the protected area. One bottle was a 1.5 ounce size with a trace amount of vanilla in the bottle, the other bottle was a 4.5 ounce size with approximately 1 ounce of vanilla. Alcohol was identified as an ingredient on the label. It was determined the vanilla extract is 35 percent alcohol by volume (ABV), above the 0.5 percent ABV considered low alcohol content. Security personnel took custody of the bottles of vanilla extract. The NRC Resident Inspector has been notified. | ||
| ENS 56830 | 3 November 2023 15:27:00 | Oconee | NRC Region 2 | B&W-L-LP | The following information was provided by the licensee via email: A press conference is being held by the Oconee County Sheriff's office today at 1530 EDT in which Duke Energy will be present and may participate regarding the facts pertaining to an event that occurred outside the Oconee Nuclear Power Plant. The NRC Resident Inspector was notified. This is a four-hour report per 10 CFR 50.72(b)(2)(xi) for any event or situation for which a news release is planned or notification to other government agencies has been or will be made which is related to heightened public or government concern. The event is not significant with respect to the health and safety of the public. | ||
| ENS 56829 | 2 November 2023 16:41:00 | Point Beach | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via email: On November 2, 2023, at 0715 CDT, it was discovered that the results of a blind performance specimen provided to a Health & Human Services (HHS)-certified testing facility were not as expected. The blind specimen results indicated a false negative for MDA/MDMA and a false positive for amphetamines. Investigation is ongoing to determine if the results are accurate. This report is being made in accordance with 10 CFR 26.719(c)(2) and 10 CFR 26.719(c)(3). The NRC Resident Inspector has been notified by the licensee. | ||
| ENS 56827 | 1 November 2023 16:52:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via email: On October 31 at 1856 CDT, Prairie Island Nuclear Generating Plant personnel identified a prohibited item (alcohol) in a kitchen area located within the protected area. An 'Extent of Condition' search was performed of all other protected area kitchen areas, no additional prohibited items were found. The NRC Resident has been notified. | ||
| ENS 56826 | 1 November 2023 09:38:00 | Hatch | NRC Region 2 | GE-4 | High Pressure Coolant Injection Reactor Protection System Reactor Core Isolation Cooling Emergency Core Cooling System Main Condenser | Manual Scram | The following information was provided by the licensee via email: At 0648 EDT on 11/1/23, with Unit 2 in MODE 1 at 56 percent power, the reactor was manually tripped due to a trip of the 'B' reactor feed pump (RFP). The 'A' RFP had been previously isolated due to a leak. Closure of containment isolation valves (CIVs) in multiple systems and the actuation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) occurred as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained with RCIC. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 was not affected. Due to the emergency core cooling system (ECCS) discharging into the reactor this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). Also, the reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, it is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs, RCIC and HPCI. There was no impact on the health and safety of the public or plant personnel. The Resident Inspector was notified. |
| ENS 56825 | 31 October 2023 14:20:00 | River Bend | NRC Region 4 | GE-6 | The following information was provided by the licensee via phone call and email: On October 31, 2023, at 0800 CDT, River Bend Station discovered that the results of a blind performance sample provided to an Health and Human Services (HHS)-certified testing facility were inaccurate (false negative). This report is being made in accordance with 10 CFR 26.719(c)(3). The HHS-certified testing facility has been informed of the error. The licensee notified the NRC resident inspector.
On November 1, 2023, River Bend Station personnel were informed by the HHS-certified testing facility that the cut-off levels used for analysis of the performance testing sample in question were the correct (higher) cut-off levels currently being used by the licensee. This resulted in a correct negative test. The performance testing sample sent to the HHS-certified testing facility was purchased for use based on the new lower cut-off levels in accordance with the new fit for duty (FFD) rule being implemented by the licensee on November 6, 2023. Because the higher confirmatory cut-off levels were used at the HHS-certified testing facility, the results provided were correct. The NRC Resident Inspector has been notified." Notified R1DO (Eve) and FFD Group (email) | ||
| ENS 56822 | 30 October 2023 17:06:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone call and email: A non-licensed supervisory employee had a confirmed positive test during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified. | ||
| ENS 56815 | 25 October 2023 15:09:00 | Pilgrim | NRC Region 1 | GE-3 | The following information was provided by the licensee via phone and email: This is a non-emergency 30-day notification for missing licensed material. This event is reportable in accordance with 10 CFR 20.2201(a)(1)(ii). On September 25, 2023, while performing the required semi-annual source leak check and inventory, radiation protection personnel could not locate seven sealed radioactive sources. Five of the sources exceed the reporting threshold of ten times the activity listed in 10 CFR 20 Appendix C. Of the five sources, four were Ni-63 sources previously utilized in security bomb detection equipment with a current source radioactivity of between 7.1 and 8.7 mCi. The fifth sealed source exceeding the reporting threshold is an Am-241 former lab calibration standard with a source radioactivity of 0.97 microcuries. These sources were last accounted for on July 6, 2022. Pilgrim's accountability process does not require leak checks or physical inventory of sources that are out of service. A search was conducted for the missing sources; however, they could not be located. These sealed sources are classified as Category 5 radioactive sources in accordance with the International Atomic Energy Agency (IAEA) Safety Guide No. RS-G-1.9. Sources that are less than Category 3 (Cat 4 and 5 sources) are very unlikely to cause permanent injury to individuals. Based on the activity of Ni-63 and Am-241 present in the sources, this 30-day phone notification to NRC is provided pursuant to 10CFR20.2201(a)(1)(ii). The required written report pursuant to 10CFR20.2201(b)(1) will be provided to NRC within 30 days. The Resident Inspector has been notified. The licensee will notify State and local authorities. THIS MATERIAL EVENT CONTAINS A 'Less than Cat 3' LEVEL OF RADIOACTIVE MATERIAL Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf | ||
| ENS 56813 | 25 October 2023 01:12:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | Reactor Protection System Auxiliary Feed Water | Automatic Scram | The following information was provided by the licensee via email: At 2159 on 10/24/2023, with Unit 3 in Mode 1 at 100 percent power, the reactor was automatically tripped due to an actuation signal into the Unit 3 reactor protection system protection rack during maintenance. The trip was uncomplicated with all systems responding normally post trip. Decay heat is being removed via auxiliary feed water system and the steam dump system. Unit 4 is not affected. This event is being reported pursuant to 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. The cause of the automatic reactor trip will be investigated by the licensee. |
| ENS 56811 | 22 October 2023 16:40:00 | Cooper | NRC Region 4 | GE-4 | Secondary containment Reactor Building Ventilation | The following information was provided by the licensee via fax and phone: On October 22, 2023, at 1149 CDT, with the reactor at 100 percent core thermal power and steady state conditions, the Cooper Nuclear Station secondary containment differential pressure exceeded the Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.1 limit of -0.25 inches water gauge. The condition existed for approximately 80 seconds until the reactor building ventilation system responded to restore differential pressure to normal. Investigations identified a hinged duct access hatch found open. The hatch was closed and latched, and ventilation system parameters were returned to normal. There were no radiological releases associated with this event. Declaring secondary containment inoperable as a result of not meeting TS SR 3.6.4.1.1 is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material and mitigate the consequences of an accident. The NRC Senior Resident Inspector has been informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At the time the licensee notified the NRC Headquarters Operations Officer, the cause of the hinged access duct being open had not been determined. This event has been added to the licensee's corrective action program. | |
| ENS 56810 | 22 October 2023 01:43:00 | Nine Mile Point | NRC Region 1 | GE-2 | Reactor Recirculation Pump | The following information was provided by the licensee via phone and email: On October 21, 2023, at 2048 EDT, reactor recirculation pump (RRP) 12 tripped. The cause for the trip is under investigation. Following the RRP trip, the average power range monitors (APRMs) flow bias trips were inoperable due to reverse flow through RRP 12. The APRMs were restored to operable on October 21, 2023, at 2058 EDT, when the RRP 12 discharge blocking valve was closed. This 8-hour non-emergency report is being made based upon requirements of 10CFR50.72(b)(3)(v)(A) which states: "Licensee shall notify the NRC of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition." The NRC Resident Inspector has been notified. | |
| ENS 56809 | 21 October 2023 09:25:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax and email: Fire potentially degrading the level of safety of the plant. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At 0907 EST, the licensee declared a notification of unusual event, under emergency action level HU.4, due to multiple fire alarms and CO2 discharge in the emergency diesel building. When the plant fire brigade entered the building, there was no indication of fire or damage to any plant equipment. The cause of the multiple alarms is under investigation. State and local authorities were notified and no offsite assistance was requested. Both units remain at 100 percent power. The NRC Resident Inspector has been notified. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
At 1007 EDT, Watts Bar terminated the notification of unusual event. The basis for termination was that no fire or damaged plant equipment was found. The NRC Resident Inspector has been notified. Notified R2DO (Miller), IR-MOC (Crouch), NRR-EO (Felts), DHS-SWO, FEMA Ops Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
Watts Bar Nuclear Plant (WBN) is retracting Event Notice 56809, Notice of Unusual Event, based on the following additional information, not available at the time of the initial notification. Specifically, in accordance with the emergency preparedness implementing procedures, WBN reported a condition that was determined to meet emergency action level (EAL) HU4, Initiating criteria number 1, receipt of multiple (more than 1) fire alarms or indicators and the fire was within any Table H2 plant area, which includes the diesel generator building. It was further determined that multiple fire detection zones actuated (spurious and invalid) enabling the discharge of installed fire suppression (CO2) into the space. Upon entry by the site fire brigade, it was determined that no smoke or fire existed and reported to the Shift Manager at 0930 EDT. All fire alarms were reset. Troubleshooting activities are in progress to determine the cause. A fire watch has been established and CO2 has been isolated. The required compensatory measures for the affected areas will remain in place until completion of the investigation, and CO2 suppression is restored to functional. Notified R2DO (Miller), IR-MOC (Crouch), NRR-EO (Felts), DHS-SWO (email), FEMA Ops Center (email), CISA Central (email), FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). | ||
| ENS 56804 | 19 October 2023 19:58:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via phone and email: Reporting due to loss of emergency preparedness capabilities. Seismic monitoring capability is non-functional due to loss of power. These monitors do not have a credited compensatory measure. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The NRC Resident Inspector has been notified. The licensee intends to notify state and local officials. | ||
| ENS 56803 | 19 October 2023 15:15:00 | Prairie Island | NRC Region 3 | Westinghouse PWR 2-Loop | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | The following information was provided by the licensee via email: On 10/19/2023, at approximately 1110 (CST), with Unit 1 in mode 1 at 100 percent power, the reactor automatically tripped. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The cause of the trip is being investigated. Operations responded and stabilized the plant. Auxiliary feedwater actuated as expected. Decay heat is being removed by the steam generator through the steam generator power operated relief valve. The trip was complex as non-safety related power was lost to both Unit 1 and Unit 2. Unit 1 is currently in mode 3 and on natural recirculation as both reactor coolant pumps are without power. Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was lost for approximately 70 minutes. No impacts to the SFP temperature were observed. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Due to the actuation of the auxiliary feedwater system following the reactor trip, this event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The second paragraph of the original report is amended as follows to correct information regarding the spent fuel pool for Unit 2: Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was maintained at all times with one train of SFP cooling. The second train lost power and was restarted approximately 70 minutes (after power was lost). No impacts to the SFP temperature were observed. Notified R3DO (Orth) and IR MOC (Crouch) and NRR EO (Felts) via email | |
| ENS 56802 | 18 October 2023 15:18:00 | McGuire | NRC Region 2 | Westinghouse PWR 4-Loop | Steam Generator Feedwater Auxiliary Feedwater | The following information was provided by the licensee via email: On October 18, 2023, at 1116 (EDT), with Unit 1 in Mode 5, an automatic actuation of the 1A auxiliary feedwater motor driven pump occurred when an incorrect action resulted in an automatic start signal. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. Feedwater is not needed for plant conditions, and the 1A auxiliary feedwater pump did not feed the steam generators. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56797 | 15 October 2023 23:30:00 | Brunswick | NRC Region 2 | GE-4 | Emergency Diesel Generator | The following information was provided by the licensee: At 2256 EDT on October 15, 2023, Brunswick declared a Notification of Unusual Event due to a fire not extinguished within 15 minutes. The licensee received fire alarms and indication of a halon discharge in the basement of the emergency diesel generator building. Due to the delay in the entry into the area, the licensee was not able to verify that the fire was out within 15 minutes. Upon entry into the room, the licensee noted an acrid odor near a transformer, but there was not a fire in the room. The fire was declared out at 2310 EDT. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
The following information was provided by the licensee via email: Termination of Unusual Event due to verification of no fire in the basement of the emergency diesel generator building." The licensee terminated the Unusual Event at 0045 on 10/16/23. The licensee notified the NRC Resident Inspector. Notified R2DO (Miller), IR-MOC (Grant), NRR-EO (Felts), DHS-SWO, FEMA Ops Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). | |
| ENS 56790 | 12 October 2023 23:31:00 | Ginna | NRC Region 1 | Westinghouse PWR 2-Loop | Reactor Protection System Control Rod Auxiliary Feed Water Main Steam | Manual Scram | The following information was provided by the licensee via email: On 10/12/23 at 2127 EDT, with the Unit 1 in Mode 1 at 100% Power, operators identified degrading condenser vacuum and manually tripped the reactor. All control rods inserted as expected. The trip was not complex, and all systems responded normally post-trip. The cause of the degraded condenser vacuum was an unexpected closure of the condenser air ejector regulator. The cause of the air ejector regulator going closed is not fully understood and is being investigated. Following the SCRAM, Operators responded and stabilized the plant. Decay heat is being removed by the Main Steam System through the Atmospheric Relief Valves (ARVs) and Auxiliary Feed Water (AFW) systems. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56787 | 11 October 2023 11:00:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | Emergency Core Cooling System | The following information was provided by the licensee via phone and email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid specific system actuation of the North Anna Power Station Unit 1 Emergency Core Cooling System (ECCS). On 6/18/2023, a comparator card power supply associated with 1-CH-PC-1121A, charging pressure low-standby pump start signal comparator, failed and caused the `A' and `B' charging pumps to auto-start and the previously running `C' charging pump to trip and lock-out. This event is considered an invalid system actuation because the actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. The ECCS pumps functioned as expected in response to the actuation. The `A' Charging pump was shut down in accordance with plant procedures following replacement of the comparator card. There was no impact on the health and safety of the public or plant personnel. The reportability requirement was determined beyond the 60-day notification requirement on 9/21/2023. The NRC Resident Inspector has been notified. | |
| ENS 56786 | 10 October 2023 19:44:00 | Cooper | NRC Region 4 | GE-4 | The following information was provided by the licensee via fax: On October 10, 2023, at 1553 CDT, Cooper Nuclear Station (CNS) was notified of a spurious actuation of a single alert notification system siren in Nemaha, Nebraska. The CNS Emergency Alert System (EAS) was not activated. The actuation occurred during siren testing conducted at approximately 1545 CDT. No emergency conditions are present at Cooper Nuclear Station. A press release from Nebraska Public Power District is not planned at this time. This condition is reportable under 10CFR 50.72(b)(2)(xi) for any event or situation for which a news release is planned or notification to other government agencies has been or will be made which is related to heightened public or government concern. The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Offsite notification was to local Nemaha County Emergency Management. | ||
| ENS 56785 | 10 October 2023 00:38:00 | Palo Verde | NRC Region 4 | CE | Reactor Coolant System | The following information was provided by the licensee via email: On October 9, 2023, during the Palo Verde Nuclear Generating Station Unit 1 refueling outage, while performing a small nozzle inspection in support of boric acid walkdowns, boric acid leakage was found on the area of the weld of a pressurizer thermowell. At 1507 MST, non-destructive examination of the weld indicated leakage through the reactor coolant pressure boundary. The exam result constitutes welding or material defects in the primary coolant system that are unacceptable under ASME Section XI. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56784 | 9 October 2023 21:52:00 | Farley | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: A non-licensed employee supervisor failed a test specified by the fitness for duty testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified. | ||
| ENS 56781 | 6 October 2023 18:12:00 | Diablo Canyon | NRC Region 4 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A). On August 8, 2023, at 1107 hours pacific daylight time (PDT) with Unit 1 in Mode 1 at 100 percent power, an invalid actuation occurred when Unit 1 4-kV vital bus 'G' was automatically transferred from auxiliary power to startup power due to an invalid bus under voltage signal, which occurred during planned maintenance activities. As a result of the actuation signal, auxiliary salt water and containment fan cooling units transferred automatically and started as designed. Plant systems responded as expected. This event was entered into the Diablo Canyon Power Plant corrective action program for resolution. There was no plant or public safety impact. The NRC Senior Resident Inspector has been notified. | ||
| ENS 56779 | 5 October 2023 12:29:00 | Vogtle | NRC Region 2 | W-AP1000 | Reactor Coolant System Automatic Depressurization System | The following information was provided by the licensee via phone and email: At 1439 EDT on August 7, 2023, a spurious level spike on the unit 4 reactor coolant system (RCS) level instrument (4-RCS-LT160A, 'Hot Leg 1 Level') caused actuation of containment isolation, reactor trip, automatic depressurization system (ADS) stage 4, and in containment refueling water storage tank (IRWST) isolation signals. The spurious level changes caused an invalid signal based on the incidental response of the 4-RCS-LT160A instrumentation due to water spray that was being used for reactor vessel cleaning (being performed prior to initial fuel loading). The level fluctuations resulted in engineered safety features actuation signals (containment isolation, ADS stage 4, and IRWST isolation signals) and a reactor trip signal, with the reactor trip signal already present. Three containment isolation valves closed due to the containment isolation signal that was generated. These valves were: 4-CAS-V014, 'instrument air supply containment isolation, air-operated valve,' 4-SFS-V034, 'spent fuel pool cooling system suction header containment isolation, motor-operated valve,' and 4-SFS-V035, 'spent fuel pool cooling system suction header containment isolation, motor-operated valve.' The other automatic containment isolation valves were either already closed at the time of the event or properly removed from service. All affected equipment functioned properly. The other actuation signals that were observed during this event (ADS stage 4, IRWST isolation, and reactor trip) did not result in any equipment changing position or automatically operating (i.e., the actuation signals occurred while the systems were properly removed from service). Units 1, 2, and 3 were not affected. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified. | |
| ENS 56774 | 3 October 2023 12:55:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | Reactor Coolant System | The following information was provided by the licensee via email: At 1154 EDT on 10/03/23, investigation into a boric acid indication was determined to be through a leak on a weld-o-let upstream of a pressurizer level transmitter isolation valve. Unit 2 is currently in MODE 6 with reactor coolant system (RCS) operational leakage limits not applicable. The leak is not quantifiable as it only consists of a small amount of dry boric acid at the location. The failure constitutes welding or material defects in the primary coolant system that are unacceptable under ASME Section XI. Therefore, this is a degraded condition reportable under 10 CFR 50.72(b)(3)(ii)(A). This condition does not affect the health and safety of the public or station employees. The Resident Inspector was notified. | |
| ENS 56769 | 1 October 2023 03:02:00 | Diablo Canyon | NRC Region 4 | Westinghouse PWR 4-Loop | Auxiliary Feedwater | Manual Scram | The following information was provided by the licensee via email: At 2014 (PDT) on 09/30/2023, with (Diablo Canyon) Unit 1 in Mode 1 at 11 percent reactor power in preparation for a pre-planned manual reactor trip into a scheduled refueling outage, the reactor was manually tripped due to a failed secondary system dump valve. Auxiliary feedwater was manually started in accordance with plant procedures. This event is being reported in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). There was no plant or public safety impact. The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Diablo Canyon Unit 2 was unaffected. |
| ENS 56759 | 27 September 2023 15:28:00 | Monticello | NRC Region 3 | GE-3 | Main Steam Isolation Valve Primary containment Reactor Pressure Vessel Main Steam Line | The following information was provided by the licensee via fax: (On 09/27/2023) at 1041 CDT, with the plant at 75 percent power and main turbine control valve testing in progress, a reactor pressure transient resulted in a reactor steam dome high pressure scram and subsequent group 1 primary containment isolation of the main steam lines (MSL). All main steam isolation valves closed as a result of the group 1 isolation signal. Additionally, a group 2 containment isolation signal was received due to reactor pressure vessel (RPV) level less than plus 9 inches during the transient. Operations personnel responded and stabilized the plant. The high-pressure coolant injection (HPCI) system was placed in service to control RPV pressure. HPCI did not inject into the RPV and was not needed to control RPV water level. The cause of the initial pressure transient is under investigation. The NRC Resident Inspector has been notified.
The utility notified the State of Minnesota and Wright and Sherburne counties. Notified R3DO (Orlikowski) | |
| ENS 56755 | 22 September 2023 22:09:00 | Turkey Point | NRC Region 2 | Westinghouse PWR 3-Loop | Auxiliary Feed Water | The following information was provided by the licensee via email: At 1819 EDT on 9/22/2023, with Unit 3 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main generator lockout. The probable cause of the main generator lockout was from a lightning strike. The trip was uncomplicated with all systems responding normally post-trip. Operations stabilized the plant in Mode 3. Decay heat is being removed by the steam dump system. Unit 4 is not affected. Auxiliary feed water was actuated as expected as a result of the reactor trip. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. | |
| ENS 56753 | 21 September 2023 10:31:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: A licensed operator had a confirmed positive test for alcohol during another entity's pre-access fitness-for-duty screening for unescorted access authorization. The individual's unescorted access at Peach Bottom Atomic Power Station has been denied. The NRC Resident Inspector has been notified. | ||
| ENS 56745 | 18 September 2023 23:26:00 | Clinton | NRC Region 3 | GE-6 | The following information was provided by the licensee via email: On 9/18/2023 at 2007 CDT, Clinton reported to the Illinois Emergency Management Agency, National Response Center and DeWitt County a hazardous substance release of 1300 gallons of Sodium Bisulphite. The release was at the site's flume discharge building due to a crack on a fitting inside the building. This release did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | ||
| ENS 56743 | 18 September 2023 14:48:00 | North Anna | NRC Region 2 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: On 09/17/2023 at 2218 (EDT), Operations identified that the bearing cooling (BC) tower basin was overflowing. Earlier in the day, the BC tower was isolated as part of a planned maintenance evolution and the overflow condition was due to isolation valve leak-by. At 2255, the leak-by was corrected and stopped the overflow. Approximately 75 gallons may have been discharged to the lake from the overflow. The BC water was sampled by Chemistry and all chemical parameters were within VP DES (Virginia Pollutant Discharge Elimination System) limits. At 1420 on 09/18/23, a 24-hour notification was made to the Virginia Department of Environmental Quality (DEQ) in accordance with the North Anna VPDES permit. This issue is being reported per 10CFR50.72 (b)(2)(xi) due to the notification of other government agency. The NRC Resident Inspector was notified. | ||
| ENS 56731 | 9 September 2023 15:35:00 | Ginna | NRC Region 1 | Westinghouse PWR 2-Loop | Reactor Protection System Control Rod Auxiliary Feed Water Main Steam | Automatic Scram | The following information was provided by the licensee via email: On 9/9/23 at 1143 EDT, with the Unit 1 in Mode 1 at 100 percent power, all 4 turbine control valves closed resulting in a reactor protection system (RPS) automatic reactor trip on over temperature differential temperature. All control rods inserted as expected. The trip was not complex and all systems responded normally post-trip. The cause of the control valve closure has not been determined. Following the SCRAM, operators responded and stabilized the plant. Decay heat is being removed by the main steam system through the atmospheric relief valves and auxiliary feed water systems. Due to the RPS actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56727 | 7 September 2023 23:35:00 | Arkansas Nuclear | NRC Region 4 | CE B&W-L-LP | The following information was provided by the licensee via email: On September 7 at 1230 CDT, Arkansas Nuclear One personnel identified 5 bottles of vanilla extract in kitchen areas located inside the Protected Area. A total of 5 bottles were identified. The bottles ranged in sizes of 1 to 4 ounces. Ingredients were listed as vanilla extracts in water and alcohol. The percentage by volume of alcohol varied from 13 - 41 percent. This report satisfied the reporting criteria of 10 CFR 26.719. The NRC Resident Inspector has been notified. |