Semantic search

This page or action requires JavaScript to work. Please enable JavaScript in your browser or use a browser where it is supported, so that functionality can be provided as requested. For further assistance, please have a look at the noscript help page.

Search Clear all entries

	Discovered date	Reporting criterion	Title	Event description
ENS 57021	11 March 2024 17:37:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Manual Reactor Trip	The following information was provided by the licensee via phone and email: On March 11, 2024, at 1337 EDT, with Unit 1 in Mode 1 at 35 percent power performing power ascension activities, the reactor was manually tripped due to the 'A' reactor feed pump (RFP) tripping on low suction pressure. Due to the power level at the time, the 'B' RFP had not been placed in service. Closure of containment isolation valves (CIVs) in multiple systems and 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. The 'B' RFP was placed in service and is controlling reactor water level. Decay heat is being removed by discharging steam to the main condenser using turbine bypass valves. Unit 2 is 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 NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the 'A' RFP is under investigation. The reactor electric plant remains in a normal lineup with both emergency diesel generators available. There were no temperature or pressure technical specification limits approached.
ENS 56997	28 February 2024 18:50:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Specified System Actuation - Automatic Start of Emergency Diesel Generators	The following information was provided by the licensee via phone and email: At 1350 EST on 2/28/2024, with Calvert Cliffs Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 65 percent power, an actuation of the '1A' and '2A' emergency diesel generators' auto-start occurred due to an undervoltage condition on the number 11 and number 21 4kV buses which are fed from the number 11 13kV bus. The '1A' and '2A' emergency diesel generators automatically started as designed when the 4kV buses' undervoltage signals were received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the '1A' and '2A' emergency diesel generators. 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 undervoltage condition was caused by the feeder breaker to the number 11 13 kV bus opening during electrical maintenance.
ENS 56993	23 December 2023 06:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - MINI-GEN Signal Generator Defect	The following is a summary of the information provided by Fairbanks Morse Engine via email: Arkansas Nuclear One (ANO) Unit 2 had a failure of a mini-gen signal generator on the opposed piston emergency diesel generator. Bench testing after removal from the engine showed an erratic signal, and this was confirmed by Fairbanks Morse. Fairbanks Morse destructive analysis revealed wear of the dynamic surface on the stator bushing inside diameter. The cause of the worn stator bushing is most likely due to inadequate lubrication on the dynamic surfaces, outside diameter of the shaft and inside diameter of the stator bushing. Possible causes of inadequate lubrication could be failure to apply enough lubrication to the dynamic surfaces during the manufacturing process or deterioration/evaporation over time. Fairbanks Morse has implemented corrective actions to address this issue, and they are estimated to be completed by May 23, 2024. Affected plants with potentially defected parts: Arkansas Nuclear One, Edwin I. Hatch Nuclear Plant, Joseph M. Farley Nuclear Generating Station, Limerick Generating Station, and Prairie Island Nuclear Generating Plant. Point of Contact: Martin Kurr Quality Assurance Manager Fairbanks Morse 608-364-8247 Martin.Kurr@fmdefense.com Fairbanks Morse Notification Report Number: 23-02
ENS 56992	22 December 2023 06:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - Air Start Valve (Bent Bottom Stem)	The following is a summary of the information provided by Fairbanks Morse Engine via email: Prairie Island Nuclear Generating Plant (PINGP) was conducting a planned replacement of emergency diesel generator air start solenoid valves when it discovered that the bottom stem appeared to be bent and observed air leakage. PINGP returned five valves to Fairbanks Morse, and they returned them to the manufacturer, ASCO. ASCO reassembled one valve and confirmed there was air leakage through the valve. The leakage path was from the air supply port to the exhaust port when the valve was in the de-energized normally open state. ASCO functionally tested the remaining four valves and found a second valve that also leaked. ASCO and Fairbanks Morse have implemented corrective actions to address this issue. Fairbanks Morse will notify PINGP and Limerick Generating Station. Affected plants with potentially defected parts: Prairie Island Nuclear Generating Plant and Limerick Generating Station. Point of Contact: Martin Kurr Quality Assurance Manager Fairbanks Morse 608-364-8247 Martin.Kurr@fmdefense.com Fairbanks Morse Notification Report Number: 23-01
ENS 56990	24 February 2024 08:19:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Start of Emergency Diesel Generator	The following information was provided by the licensee via phone and email: At 0219 CST on February 24, 2024, Browns Ferry Unit 3 was shut down in a refueling outage, while closing 4 kV shutdown board breaker 3EB-9, the 4 kV shutdown board normal feeder breaker tripped open resulting in a valid 4 kV bus under-voltage condition. Due to the under-voltage condition, the 3B emergency diesel generator (EDG) auto started and tied to the board. The cause of the breaker tripping open is unknown and an investigation is in progress. All systems responded as expected for the loss of voltage. This event requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A). There was no impact to 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 safety related equipment was affected. The 3B EDG continues to supply the shutdown board pending further investigation.
ENS 56988	28 December 2023 13:15:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Emergency Diesel Generators	The following information was provided by the licensee via phone and email: This 60-day optional telephone notification is being made in lieu of a Licensee Event Report (LER) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0815 EST on December 28, 2023, an invalid actuation of the four emergency diesel generators (EDGs) occurred. It was determined that this condition was likely caused by spurious operation of the undervoltage relay for the startup auxiliary transformer feeder breaker to the `1D' balance of plant bus which was being fed by the unit auxiliary transformer at the time, per the normal lineup. This non-safety related EDG actuation logic was disabled, and additional investigation is planned during the upcoming refueling outage. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. During this event, the four EDGs functioned successfully, and the actuations were complete. All emergency buses remained energized from offsite power and, therefore, the EDGs did not tie to their respective buses. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector had been notified.
ENS 56978	19 February 2024 07:36:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Start of 'B' Emergency Diesel Generator	The following information was provided by the licensee via phone and email: On February 19, 2024, at 0236 EST, with VC Summer Unit 1 in Mode 1 at 100 percent power, an actuation of the `B' emergency diesel generator (EDG) occurred. The reason for the `B' EDG auto-start was the trip of 1 `DB' normal incoming breaker. The `B' EDG automatically started as designed when the undervoltage signal was received. The `B' emergency feedwater pump started due to the undervoltage signal and ran for approximately 1 minute and was secured by operations per procedure. Other plant equipment and systems also responded as expected. 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 `B' EDG and a valid actuation of the `B' emergency feedwater pump. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The `A' Emergency Diesel Generator was tagged out for maintenance earlier in the shift, but maintenance has not started. The plan is to restore the `A' emergency diesel generator to an operable status and investigate the cause of the 1 `DB' normal incoming breaker trip. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: This event resulted in the plant entering a 12 hour limiting condition for operation (LCO) in accordance with technical specification (TS) 3.8.1.1.C. due to having one operable EDG and a loss of offsite power.
ENS 56977	19 February 2024 04:25:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Start of Emergency Diesel Generator	The following information was provided by the licensee via phone and email: At approximately 2325 EST on February 18, 2024, with Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 100 percent power, emergency diesel generator 2 automatically started due to the unexpected loss of AC power to emergency bus E2 during a planned transfer of E2 DC control power from normal to alternate for the 1B-1 battery. In addition, the unexpected loss of AC power to E2 resulted in Unit 1 primary containment isolation system (PCIS) partial Group 2 (i.e., drywell equipment and floor drain, residual heat removal (RHR), discharge to radioactive waste, and RHR process sample), Group 6 (i.e., containment atmosphere control/dilution, containment atmosphere monitoring, and post accident sampling systems), and partial Group 10 (i.e., air isolation to the drywell) isolations. Emergency diesel generator 2 automatically started and re-energized the E2 bus as designed when the loss of E2 signal was received. The PCIS actuations were as expected for the outage plant line up on Unit 1 at the time. The cause of the loss of electrical power to emergency bus E2 is under investigation at this time. 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 emergency diesel generator 2 and PCIS. 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: This event will be entered into the plant's corrective action program.
ENS 56970	16 February 2024 03:24:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Actuation of Emergency Diesel Generator System	The following information was provided by the licensee via email: At 2224 EST on February 15, 2024, with both units 1 and 2 in mode 1 at 100 percent power, an actuation of the emergency diesel generator (EDG) system on 1A-A, 1B-B, and 2B-B EDGs occurred while removing clearances. The 2A-A EDG did not start because it was still under a clearance. The reason for the emergency diesel generator system auto-start was clearance removal sequencing errors. The emergency diesel generator system automatically started as designed when the common emergency start signal was received. 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 emergency diesel generator system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. * * RETRACTION ON 2/21/2024 AT 1549 EST FROM TYSON JONES TO KAREN COTTON * * * The following information was provided by the licensee via email: In accordance with NUREG-1022, Section 2.8 and Section 4.2.3, Watts Barr is retracting the previous report EN 56970 pursuant to 10 CFR 50.72(b)(3)(iv)(A). The start signal for the 1A-A, 1B-B, and 2B-B emergency diesel generators (EDG)s was from activation of the common emergency start of the 2A-A EDG. The actuation was not from a loss of offsite power (LOOP) to any shutdown board or from any parameters that would initiate a safety injection (SI) signal, for which the EDG is designed to provide a design basis safety function. Also, the starts were not from intentional manual actuation. Starting the EDGs did not make them inoperable and each EDG was able to perform its design safety function. The common emergency start relay for each diesel is not safety related. It is an anticipatory and redundant circuit to start other EDGs in the event of a LOOP or SI related to the specific EDG. With the 2A-A EDG out of service, the associated common emergency circuit would not be required to perform any function. The starts were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the system. Since the starts were not initiated via an automatic signal from a LOOP, SI, or traditional operator action, the signal is not a valid actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). Therefore, EN 56970 is being retracted. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Miller)
ENS 56920	5 May 2021 05:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - Gould Field Flash Relay on Emergency Diesel Failed	The following information was provided by the licensee via email and phone call: On January 9, 2024, VC Summer Nuclear Station (VCSNS) determined a manufacturing defect affecting a field flash contactor (K2) relay on its 'B' emergency diesel generator (EDG) was reportable under Part 21. During testing on May 5, 2021, the 'B' EDG was rendered inoperable when its K2 relay coil switch exhibited intermittent binding due to insufficient clearances of the switch actuator from the protective case and plastic switch molding. The inadequate clearances resulted in accelerated loss of graphite lubrication at these locations, which led to mechanical binding. VCSNS replaced the affected relay and restored operability of its 'B' EDG. Manufacturer/Model: Gould F10NOCLD1 DC coil switch A written notification in accordance with 10 CFR 21.21(d)(3)(ii) will be provided within 30 days. 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: The licensee did not notify the intellectual property owner because the manufacturer is no longer operating or in business. They are not aware of other plants that utilize this coil switch.
ENS 56919	12 November 2021 05:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - EATON-CUTLER Hammer Relay on Emergency Diesel Failed	The following information was provided by the licensee via email and phone call: On January 9, 2024, VC Summer Nuclear Station (VCSNS) determined a manufacturing defect affecting a control power circuit monitor (CP1) relay on its 'B' emergency diesel generator (EDG) was reportable under Part 21. On November 12, 2021, the 'B' EDG was rendered inoperable when its CP1 relay de-energized due to mechanical binding of the magnet carrier assembly. The binding was caused by a manufacturing defect that allowed heat-induced shrinkage to reduce the clearance between the magnet carrier and adjacent coil housing and base, preventing it from moving freely. VCSNS replaced the affected relay and restored operability of its 'B' EDG. Manufacturer/Model: Eaton-Cutler Hammer D26MRD30A1 A written notification in accordance with 10 CFR 21.21(d)(3)(ii) will be provided within 30 days. 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: The plant has notified the manufacturer. It is not known if any other plants are affected by this defect.
ENS 56913	27 November 2023 05:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - EMD Cylinder Liner with Bore Deficiency	The following is a summary of the information provided by Engine Systems, Inc. (ESI) via fax: An edge or lip in the bore of an EMD (Brand name: Electro-Motive Diesel) cylinder liner prevented successful installation of the corresponding power assembly on an emergency diesel generator set. The lip is located axially at the bottom of the inlet ports and is present around the circumference of the bore. The EMD model 645E4 is a 2-stroke engine with air inlet ports in the wall of the cylinder liner. As the piston travels below the inlet ports, air box pressure scavenges and replenishes air to the power assembly. Installation of the power assembly requires lowering the piston through the liner in order to secure the connecting rod to the crankshaft. During this process the piston could not be lowered below the inlet ports due to the piston rings catching on the lip. The power assembly was not installed and therefore there was no safety hazard; however, if the defect had gone undetected there was the potential to damage engine components and possibly reduce load carrying capacity of the engine. The extent of the condition is this single cylinder liner, P/N 9318833, S/N 20M0938 used in the power assembly at Tennessee Valley Authority (TVA) - Sequoyah Nuclear Plant, Serial Number: 23H1306. Corrective Actions: For TVA-Sequoyah: No action required; the power assembly has been returned to ESI. For ESI: To prevent reoccurrence, ESI has revised the dedication package to include verification that bore machining is continuous along the entire length and no edges or lips are present. The revision was implemented on December 6, 2023. * * UPDATE ON 1/17/24 AT 1515 EST FROM DAN ROBERTS TO ADAM KOZIOL * * * Engine Systems, Inc. sent a revision to change the date of defect identification to November 27, 2023. Notified R2DO (Miller), Part 21 Group (email)
ENS 56905	21 December 2023 16:23:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - Fuel Oil Piping Defect	The following summary was provided by the licensee via phone and email: On December 21, 2023, VC Summer determined that the original equipment manufacturer of its Emergency Diesel Generator (EDG) did not provide adequate documentation for an exception to ASME Code, Section III, Class 3 for the on-skid EDG fuel oil system. Specifically, the on-skid fuel oil threaded piping design used a code-allowed exception without providing adequate evaluation documentation supporting threaded Schedule 40 piping. On November 2, 2022, the 'A' EDG fuel oil piping failed during routine testing, requiring it to be declared inoperable until repairs were implemented. The NRC Senior Resident Inspector has been notified. Corrective actions: VC Summer has compensatory measures in place and is scheduled to implement a modification during the first quarter of 2024. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: In addition to VC Summer, it was discussed via phone that Farley has a similar design for its on-skid EDG fuel oil system and potentially might be affected by this Part 21 report. No other sites have been discussed at this time. The EDG manufacturer is: Fairbanks-Morse Corp. EDG Model: Colt-Pielstick 12-Cylinder PC-2.2 VC Summer POC for additional information: Justin Bouknight, Licensing Engineer, (803) 941-9828, justin.bouknight@dominionenergy.com.
ENS 56877	2 December 2023 12:10:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Specified System Actuation - Auto Start of Emergency Diesel Generator	The following information was provided by the licensee via email: At 0610 CST on 12/2/2023, with Unit 2 in Mode 1 at 100 percent power, the South Texas Project switchyard south electrical bus was de-energized. Emergency diesel generator (EDG) '22' automatically started in response to the loss of offsite power on the train 'B' engineered safety feature (ESF) electrical bus. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of an emergency AC electrical power system (50.72(b)(3)(iv)(B)(8)). All required loads were successfully started. 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 initial loss of the south electrical bus, partial loss of off-site power, put the plant in a 24 hour limiting condition for operation (LCO) in accordance with (IAW) technical specification (TS) 3.8.1.1.E. Power was restored to the train 'B' ESF bus via an alternate offsite power source and the EDG was returned to its automatic standby condition. Currently, the plant is in a 72 hour LCO IAW TS 3.8.1.1.A.
ENS 56863	18 November 2023 05:55:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram	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 21:41:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Essential Chiller Trains Inoperable	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 56797	16 October 2023 02:56:00	10 CFR 50.72(a)(1)(i), Emergency Class Declaration	Unusual Event Due to Fire Not Verified to Be Extinguished within 15 Minutes	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). * * UPDATE AT 0047 EDT ON 10/16/2023 FROM JOSEPH STRNAD TO BILL GOTT * * * 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 56789	11 August 2023 04:00:00	10 CFR 21.21(a)(2), Interim Report for Comply or Defect in Component	Part 21 Report - Emergency Diesel Generator Digital Reference Unit Problem	The following information was provided by the licensee via email: Donald C. Cook Nuclear Power Plant completed an internal Part 21 evaluation concerning an issue with an Emergency Diesel Generator (EDG) Digital Reference Unit (DRU) supplied by Engine Systems Incorporated (Appendix B Supplier for Woodward Governors). (On August 8, 2023,) a potential defect was identified (during a surveillance test) concerning a marginal solder joint on the DRU electronic circuit board that can result in a loss of continuity between the termination strip and the electronic board, causing a loss of setpoint output from the DRU to the Electronic Governor, and a subsequent loss of fuel to the EDG and inability to support any load. A formal failure analysis is ongoing at the time of this notification. A written notification will be provided within 30 days. Affected known plants include only Donald C. Cook Nuclear Power Plant Units 1 and 2 at the time of notification. The NRC Resident has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The EDG DRU was replaced after discovery of the potential defect and the EDG is currently operable.
ENS 56603	17 May 2023 04:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - Emergency Diesel Generator - Incorrect Bus BAR Installation	The following is a summary of the information provided by Engine Systems, Inc. (ESI) via fax: Copper bus bars are used at rectifier diode pigtail connections CR1, CR2, CR3, CR4, and CR7 to provide a common connection point for associated wiring to minimize the number of conductors on a single fastener. It is intended for the conductors to clamp directly to the bus bar, thereby maintaining a low resistance conduction path between components. ESI has determined the bus bars were installed incorrectly whereby the conductors are not clamped in direct contact, resulting in current now passing through the stainless steel fasteners. This configuration is undesired and will result in unnecessary heat generation which may lead to failure of the connection and therefore failure of the automatic voltage regulator (AVR) assembly. The AVR is relied upon to automatically regulate emergency diesel generator (EDG) terminal voltage. Failure of the AVR would impact the ability of the EDG to perform its safety-related function and therefore may impact the nuclear plant's ability to manage safety-related loads during an emergency event. This Part 21 applies to the bus bar installation for part numbers 72-12300-100-ESI, 72-14200-100-ESI, and 72-14000-100-6020. These part numbers impact Constellation Energy - Nine Mile Point, Avaltec/CFE - Laguna Verde (Div III), and Avaltec/CFE - Laguna Verde (Div II), respectively. Corrective Actions: Voltage regulator installed: Bus bar installation should be corrected immediately by restacking the components to ensure direct contact with all conductors. Voltage regulators in inventory (not installed): The bus bar installation should be corrected prior to installation. If desired, the customer may perform on site or the voltage regulator chassis may be returned to ESI for rework.
ENS 56570	13 June 2023 03:33:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Accident Mitigation - High Pressure Coolant Injection Declared Inoperable	The following information was provided by the licensee via email: At 2333 EDT on June 12, 2023, the division 2 Mechanical Draft Cooling Tower (MDCT) Fan `D' was declared inoperable due to a trip of the fan while running in high speed. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the division 2 Emergency Equipment Cooling Water (EECW) system. The EECW system cools various safety related components, including the High-Pressure Coolant Injection (HPCI) system room cooler. An unplanned HPCI inoperability occurred based on a loss of the HPCI room cooler. The cause of MDCT Fan `D' trip is currently unknown with trouble shooting being developed for remediation of the condition. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. The NRC Senior Resident Inspector has been notified. * * RETRACTION AT 1540 EDT ON 8/8/2023 FROM WHITNEY HEMINGWAY TO BILL GOTT * * * The purpose of this notification is to retract a previous event notification (EN) 56570 reported on June 13, 2023, at 0602 EDT. The cause of the fan trip was a failed vibration switch. At 0429 EDT on June 14, 2023, the vibration switch was replaced, the MDCT fan "D" was tested satisfactory for operability, and the UHS, emergency diesel generator 13/14, and MDCT were declared operable. Following the initial EN, further analysis of the condition was performed utilizing a previously performed gothic analysis model (to perform HPCI room heat-up calculations) which bounded this condition. Based on the initial conditions at the time of the indication loss, specifically HPCI room and suppression pool temperature, it was determined that the resulting worst case post-accident room temperature was sufficiently low enough to provide margin to HPCI operability without the room cooler in service for the required mission time. No other concerns were noted during the event. HPCI remained operable and there was no loss of safety function. The fan trip did not involve a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident under 10 CFR 50.72(b)(3)(v)(D). Therefore, the NRC non-emergency 10CFR50.72(b)(3)(v)(D) report was not required and the NRC report 56570 can be retracted, and no licensee event report under 10 CFR 50.73(a)(2)(v)(D) is required to be submitted. The licensee notified the NRC Resident Inspector. Notified R3DO (Nguyen)
ENS 56564	12 April 2023 15:07:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Actuation of "B" Emergency Diesel Generator Emergency Power Sequencer	The following information was provided by the licensee via email: On April 12, 2023, with Seabrook Station Unit 1 in Mode 6 at zero percent power, a valid actuation of the 'B' emergency diesel generator (EDG) emergency power sequencer occurred due to a loss of power to the 'B' train emergency bus. The 'B' EDG was removed from service for scheduled maintenance during this time. This event is being reported pursuant to 10 CFR 50.72(b)(3)(iv)(A) for a valid actuation of the 'B' EDG emergency power sequencer. The NRC Resident Inspector has been notified.
ENS 56551	2 June 2023 11:05:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Trip	The following information was provided by the licensee via email: The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS (Emergency Notification System) or under the reporting requirements of 10 CFR 50.73. At 0405 MDT on June 2, 2023, the Unit 2 reactor automatically tripped on low steam generator water levels due to degraded flow from the A main feedwater pump. Steam generator water levels reached the automatic Auxiliary Feedwater Actuation System (AFAS) setpoint resulting in automatic AFAS-1 and AFAS-2 actuations and subsequent start of both class auxiliary feedwater pumps. Steam Generator water levels are being restored to normal band with the class 1E powered motor driven auxiliary feedwater pump. Following the reactor trip, all control element assemblies inserted fully into the core. No emergency plan classification was required per the Emergency Plan. Safety related buses remained powered from offsite power during the event and the offsite power grid is stable. Both emergency diesel generators automatically started on the AFAS-1 and AFAS-2 actuations as designed and are currently running unloaded. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and a specified system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been informed. Unit 1 and 3 are in Mode 1 at 100 percent power. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Decay heat is being removed to main condenser via automatic steam bypass and B auxiliary feedwater pump.
ENS 56541	25 May 2023 17:45:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition of Emergency Diesel Generator	The following information was provided by the licensee via phone and email: At 1345 EDT on May 25, 2023, it was determined that a fire barrier for area 737-A1B was not installed, and would render the 2A Emergency Diesel Generator (EDG) not operable in the event of a fire on the Unit 2 side of elevation 737 in the Auxiliary Building. The 2A EDG is the credited power source for fire safe shutdown for a fire located in this area. Without the credited source of power, this places WBN U2 (Watts Bar Nuclear Unit 2) in an unanalyzed condition. A fire watch has been established in the area until the issue is resolved. Therefore, this event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
ENS 56538	28 March 2023 04:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - Defect Identified in Emergency Diesel Generator Governor	The following is a synopsis of information provided by the Engine Systems, Inc (ESI) via fax: Component Description: Woodward Governor, Part No. 9903-722, Serial No. 18847017 Problem Description: An EGB-35P governor/actuator (governor) installed on a customer's emergency diesel generator failed soon after installation. Investigation revealed a piece of foreign material, a loose buffer plug, inside the governor that caused the failure. Since the governor is used to maintain fuel rack position of the diesel engine, failure of the governor would prevent the emergency diesel generator from performing its safety-related function during an event. Affected Plants: Brunswick Nuclear Plant Corrective Actions for Brunswick Nuclear Plant: No action required. The affected governor has been returned to ESI. Corrective Actions for ESI: The governor will be refurbished under ESI's 10 CFR 50 Appendix B program and certified for continued use at Brunswick Nuclear Plant. To prevent reoccurrence, ESI will revise the dedication requirements to enhance existing foreign material inspection practices to include a visual inspection where the buffer plug was located within the governor. The revisions are expected to be complete within 30 days but in all cases prior to future shipments.
ENS 56430	23 March 2023 21:36:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Emergency Diesel Generator Actuation	The following information was provided by the licensee via email: At 1736 EDT on March 23, 2023, during overcurrent testing of the '2B' (Emergency Safeguards System) ESS Bus, the work group was re-installing tested relays and inadvertently caused a '2B' ESS Bus lockout. This resulted in the '2B' ESS Bus deenergizing and a valid start signal provided to the 'B' Emergency Diesel Generator (EDG). The 'B' EDG started and functioned as designed. This is being reported as an unplanned actuation of systems that mitigate the consequences of significant events in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified.
ENS 56347	25 December 2022 11:37:00	10 CFR 50.73(a)(1), Submit an LER	60-DAY Optional Telephonic Notification of Invalid Actuation of Emergency Diesel Generator (EDG)	The following information was provided by the licensee via email: This 60-day optional telephone notification is being made in lieu of an LER (Licensee Event Report) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0637 EST on December 25, 2022, the 2B EDG inadvertently started and ran unloaded without a valid undervoltage or safety injection actuation signal. It was determined that this condition was caused by the failure of the emergency start button due to age-related degradation. The button is normally held depressed (closed) by the glass enclosure in standby. To start the EDG using the Emergency Start Button, the button is released (open) when the glass enclosure is broken, which sends a start signal to the EDG. During troubleshooting, the resistance across the button contacts was measured at zero volts DC, indicating the button had failed to an open state causing the EDG to start. The button fell apart when the glass enclosure was removed. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the emergency diesel generator. The NRC Resident Inspector has been notified.
ENS 56294	4 January 2023 03:59:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Loss of Offsite Power Source	The following information was provided by the licensee via email: At 2259 EST on 1/3/2023, with Unit 1 and Unit 2 in Mode 1 at 100 percent power, an actuation of the Unit 1 B and Unit 2 A emergency diesel generator (EDG) systems, as well as an actuation of the associated auxiliary feedwater (AFW) systems on each unit occurred. The reason for the EDG auto-starts was due to a loss of an offsite power source (loss of one of the two reserve auxiliary transformers (RAT) on each unit) to the Unit 1 B and Unit 2 A safety related buses. The EDG and AFW systems automatically started as designed when the valid undervoltage signal on the affected safety related bus was received. This event is 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 Emergency Diesel Generator and the Auxiliary Feedwater Systems for both Unit 1 and Unit 2. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 56283	2 November 2022 23:29:00	10 CFR 50.73(a)(1), Submit an LER	60-DAY Telephonic Notification - Invalid Specific System Actuation	The following information was provided by the licensee via email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid specific system actuation of the Emergency Service Water System (ESW). On 11/2/2022, during normal reactor operations, multiple main control room alarms were received for D12 Emergency Diesel Generator (EDG) running and Unit 1 Division 2 Safeguard Battery Ground. The D12 EDG did not start; however, the 'B' ESW Pump auto started. Subsequent troubleshooting determined that the cause of the D12 EDG running alarms and the inadvertent auto start of the 'B' ESW Pump was a malfunction on the D12 EDG speed switch. This event is considered an invalid system actuation because the 'B' ESW Pump started in response to a false signal that the D12 EDG was running when D12 EDG did not start. This was a complete actuation of the ESW System and the system functioned as expected in response to the actuation. The affected ESW Pump was shut down in accordance with plant procedures and the degraded D12 EDG speed switch was replaced. There was no impact on the health and safety of the public or plant personnel. The licensee notified the NRC Resident Inspector.
ENS 56200	27 October 2022 05:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 Report - Fairbanks Morse Engine Emerency Diesel Generator Electronic Speed Control Module	The following information was provided by Fairbanks Morse (FM) via email: Facilities affected: FM Sales Order Item Affected Facility Serial Number 40135890 12996949 Farley 22436624 40132483 12996949 Limerick 22444358 40130845 12996949 Limerick 22444359 40130158 12996949 Hope Creek 22277182 Basic component which fails to comply or contains a defect: Emergency Diesel Generator Electronic Speed Control Module, Woodward 2301A Nature of defect: In some 2301A controls, 1 nF capacitors may have been loaded in place of 150 pF capacitors. This could affect circuitry controlling the units' crystal, speed signal and reset dynamics, and power supply operation. Safety hazard which could be created by such defect: This issue can prevent affected 2301A controls from starting up or may lead to unscheduled shutdown of affected controls. It can also prevent the prime mover from obtaining a stable speed response, causing it to hunt or overspeed. In some cases, the RESET potentiometer may run out of range to adjust the unit for stable operation or desired prime mover speed response, resulting in prime mover performance outside of specification limits. Fairbanks Morse Engine will notify affected licensees no later than 3 Nov 2022, and repair returned affected units. Additional corrective actions will be documented in the Fairbanks Morse corrective system under PD-1102. Any installed affected Speed Control should be removed from service as soon as practical and returned to Fairbanks Morse for repair. If affected controller is installed and licensee experiences unstable speed response, the electronic speed control should be turned off and the emergency diesel generator should be allowed to operate using the mechanical governor system. POC: Martin Kurr, Quality Assurance Manager (608) 364-8247
ENS 56044	15 August 2022 00:02:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Actuations of Unit 1 and Unit 3 Emergency Diesel Generators and Unit 1 Auxiliary Feedwater Pump	The following information was provided by the licensee via phone and email: At 1702 MST on August 14, 2022, a start-up transformer de-energized, resulting in a loss of power to the Unit 1 Train B 4.16 kV Class 1E Bus and the Unit 3 Train A 4.16 kV Class 1E Bus. The Unit 1 Train B Emergency Diesel Generator (EDG) and Unit 3 Train A EDG automatically started and energized their respective 4.16 kV Class 1E Buses. As a result of the loss of power on the Unit 1 Train B 4.16 kV Class 1E Bus, the B Auxiliary Feedwater Pump automatically started, as expected. The B Auxiliary Feedwater Pump was not needed for steam generator level control and no auxiliary feedwater valves repositioned. The B Auxiliary Feedwater Pump did not supply feedwater to the steam generators. 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 emergency AC electrical power systems and an auxiliary feedwater system. All systems operated as expected. Per the Emergency Plan, no classification was required due to the event. The 4.16 kV Class 1E Buses in Unit 2 were not affected by the de-energization of the start-up transformer since it was not aligned as normal power for Unit 2. Units 1, 2 and 3 remain in Mode 1 at 100% power. The cause of the start-up transformer being de-energized is under investigation. No plant transient occurred as a result of this failure. The NRC Resident Inspectors have been informed.
ENS 55878	5 May 2022 02:55:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Actuation of Unit 2 and Unit 3 Emergency Diesel Generators and Unit 3 Auxiliary Feedwater Pump	The following information was provided by the licensee via email: At 1955 on May 4, 2022, a start-up transformer de-energized, resulting in a loss of power to the Unit 2 Train A 4.16 kV Class 1E Bus and the Unit 3 Train B 4.16 kV Class 1E Bus. The Unit 2 Train A Emergency Diesel Generator (EDG) and Unit 3 Train B EDG automatically started and energized their respective 4.16 kV Class 1E Buses. As a result of the Loss of Power on the Unit 3 Train B 4.16 kV Class 1E Bus, the B Auxiliary Feedwater Pump automatically started, as expected. The B Auxiliary Feedwater Pump was not needed for steam generator level control and no auxiliary feedwater valves repositioned. The B Auxiliary Feedwater Pump did not supply feedwater to the steam generators. 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 emergency AC electrical power systems and an auxiliary feedwater system.
ENS 55859	7 March 2022 04:40:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation 60-DAY Telephone Notification	The following information was provided by the licensee via fax or email: This 60-day telephone notification is being made in lieu of an LER submittal per 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for invalid actuations of systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0040 Eastern Standard Time (EST) on March 7, 2022, Unit 1 received inadvertent High-Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiation signals. Subsequently, at approximately 0148 EST on March 7, 2022, Unit 1 received inadvertent Low-Pressure Coolant Injection (LPCI) and Core Spray initiation signals. In addition, all four Emergency Diesel Generators auto started, Group 10 (Instrument Air) Primary Containment Isolation System actuations occurred, and the Residual Heat Removal (RHR) Service Water Booster pumps tripped resulting in a brief interruption (approximately 9 minutes) to the Shutdown Cooling (SDC) heatsink. Jumpers, installed per planned refueling outage activities, prevented discharge of Emergency Core Cooling Systems into the reactor. HPCI, RCIC, and RHR Loop `A' were removed from service and under clearance. RHR SDC remained operable via RHR Loop `B' and forced circulation was maintained in the reactor. At the time of these events, Unit 1 was shutdown for refueling and the `A' and `C' reactor water level transmitters had been isolated in preparation for planned replacement. Leak-by of the instrument isolation valves occurred on both transmitters. Leak-by on the `C' instrument occurred at a faster rate with the `A' instrument providing the confirmatory signals resulting in Low Level 2 (LL2) and Low Level 3 (LL3) indication at approximately 0040 EST and 0148 EST, respectively. All actuations occurred as designed for LL2 and LL3 signals. During these events, reactor water level remained stable at the Reactor Vessel Head Flange and the `B' and `D' reactor water level transmitters remained off-scale-high, as expected under these conditions. Therefore, the actuations were not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system (i.e., there was no low reactor water level condition). Considering the above, these actuations were invalid. There was no impact on the health and safety of the public or plant personnel.
ENS 55807	27 January 2022 04:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - ELECTRO-MOTIVE Diesel (EMD) Cylinder Head with Fireface Thickness Below Specification	The following is a summary of the information provided by Engine Systems, Inc. (ESI) via fax: ESI reported that a fatigue crack was discovered in the fireface of an EMD cylinder head, P/N 40121485, D/C 18K, installed on an emergency diesel generator set. The crack initiated on the coolant side and propagated through the fireface wall of the combustion side resulting in a water leak. Fatigue failure was likely caused by a reduced fireface thickness which reduced overall rigidity of the fireface, allowing increased deformation and ultimately failure due to high tensile stress at the blend between the fireface and valve seat. A reduced fireface thickness could result in a through wall crack that would introduce jacket water into the combustion chamber. Over time if the crack propagated or went undetected engine damage may occur. Ultimately, a crack in the fireface could lead to a failure of the diesel engine which would prevent the emergency diesel generator set from performing during a safety-event. This Part 21 applies to the Cylinder heads from D/C 18K, supplied within some power pack assemblies at Watts Bar Nuclear Power Plant. Corrective Actions: ESI recommends an ultrasonic thickness inspection on the fireface to confirm thickness is within specified range to plants with these power pack assemblies. ESI has also revised its dedication package to increase the number of ultrasonic thickness inspection points. An additional enhancement is the inclusion of an inspection map for guidance and clarity of the locations to be measured. The revision was implemented on March 15, 2022. Technical questions concerning this notification can be directed to Dan Roberts, Quality Manager and John Kriesel, Engineering Manager at (252) 977-2720.
ENS 55692	6 January 2022 12:03:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Specified System Actuation - Auto Start Emergency Diesel Generator	At 0603 CST on 1/6/2022, with Unit 2 in Mode 1 at 100 percent power, the South Texas Project (STP) south switchyard electrical bus was de-energized momentarily and re-energized approximately 40 seconds later. Emergency Diesel Generators (EDG) 22 automatically started in response to loss of offsite power on Train B Engineered Safety Feature (ESF) Bus. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of an emergency AC electrical power system (50.72(b)(3)(iv)(B)(8)). 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: Unit 2 is in a 72 hour LCO per TS 3.8.1.1.A for the loss of one offsite power supply. The plant is in a normal electrical lineup. There was no impact on Unit 1.
ENS 55627	6 December 2021 16:25:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Safety System Actuation	On December 6, 2021, at 1125 hours Eastern Standard Time (EST), during planned maintenance activities, electrical power was lost to the 4160V emergency bus E-3. The power loss to emergency bus E-3 affected both Unit 1 and 2. Emergency Diesel Generator #3 received an automatic start signal but was under clearance for planned maintenance. Emergency bus E-3 was re-energized at 1315 EST hours via offsite power. The loss of power to E3 resulted in Unit 2 Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 3 (i.e., Reactor Water Cleanup), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., air isolation to the drywell) isolations. The actuations of PCIVs were completed and the affected equipment responded as designed. Per design, no Unit 1 safety system group isolations or actuations occurred. Other Unit 2 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start signal to the Standby Gas Treatment (SGT) System trains A and B and the Control Room Emergency Ventilation System (CREV). Systems functioned as designed. Safety systems functioned as designed following the de-energization of bus E-3. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of PCIS on Unit 2 and an automatic start signal to Emergency Diesel Generator #3. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Except for the Emergency Diesel Generator, which is out of service for planned maintenance, all equipment has been returned to its normal alignment. * * UPDATE FROM JJ STRNAD TO THOMAS KENDZIA AT 2028 EST ON DECEMBER 6, 2021 * * * The loss of power to E3 resulted in Unit 1 Primary Containment Isolation System (PCIS) Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems). Other Unit 1 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start signal to the Standby Gas Treatment (SGT) System trains A and B and the Control Room Emergency Ventilation System (CREV). Systems functioned as designed. Safety systems functioned as designed following the de-energization of bus E-3. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of PCIS on Unit 1. All Unit 1 equipment was returned to its normal alignment. The NRC Resident will be notified. Notified R2DO (Miller).
ENS 55605	9 November 2021 17:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 Report - Hinge Pin Retainer Plug Defective on Diesel Check Valve	The following is a synopsis of information received via facsimile: The hinge pin retainer plug used on an emergency diesel generator (EDG) stainless steel check valve exhibited low breakaway torque and thus minimal resistance to loosening when subjected to engine operating vibrations. The EDG check valve is used specifically for lube oil (LO) applications in the gallery fill line between the LO cooler and main engine pressure pump discharge elbow. Consequently, if the plug were to completely dislodge, followed by the associated hinge pin, the pressure boundary of the LO system would be compromised and oil would discharge through the 3/16 inch opening. The dedication procedure for this check valve is currently undergoing revision to incorporate a rework activity that will eliminate unintended plug loosening from future shipments. The vendor expects this to be completed by December 7, 2021. The potentially affected components were shipped to the following plants: Beaver Valley, Browns Ferry, Dresden, and Surry. Technical questions concerning this notification can be directed to Dan Roberts, Quality Manager and John Kriesel, Engineering Manager at (252) 977-2720.
ENS 55602	30 September 2021 14:07:00	10 CFR 50.73(a)(1), Submit an LER	60-DAY Optional Telephonic Notification for an Invalid Emergency Diesel Generator Actuation	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. At 0907 (EDT) on September 30, 2021, with Unit 1 in Mode 1, at 100 percent power, an actuation of the 1-1 emergency diesel generator (EDG) occurred during loss of voltage relay functional testing. The 1-1 EDG auto-start was due to human error during performance of the test procedure when the bus 1AE undervoltage signal was improperly defeated and a simulated undervoltage signal was applied. No actual undervoltage condition was present during this event. The 1-1 EDG automatically started as designed when the bus undervoltage signal was received. This was a complete actuation of an EDG to start and come to rated speed, and all affected systems functioned as expected in response to the actuation. Following the actuation, the relays were restored and the 1-1 EDG was shut down in accordance with plant procedures. This event is considered an invalid system actuation reportable under 10 CFR 50.73(a)(2)(iv)(A). The actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. Therefore, in accordance with 10 CFR 50.73(a)(1), this telephone notification is provided within 60 days after discovery of the event instead of submitting a written Licensee Event Report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55549	27 October 2021 18:29:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Actuation of the 1B EDG Load Sequencer on Bus Undervoltage Condition	At 1429 EDT on October 27, 2021 with Unit 1 in Mode 6 at 0 percent power, the 1B Emergency Diesel Generator (EDG) Load Sequencer was actuated by a valid undervoltage condition on the 1B 4160V Essential Bus that occurred during 1B Sequencer calibration activities. Valid signals were sent to both the 1B EDG and Unit 1 Auxiliary Feedwater (CA) systems. Neither system automatically started as they were both removed from service for maintenance activities at the time. 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 1B EDG and Unit 1 CA systems. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55534	20 October 2021 21:46:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Specified System Actuations of Unit 1 and Unit 3 Emergency Diesel Generators	At 1446 MST on October 20, 2021, a start-up transformer de-energized, resulting in a loss of power to the Unit 1 Train B 4.16 kV Class 1E Bus and the Unit 3 Train A 4.16 kV Class 1E Bus. The Unit 1 Train B Emergency Diesel Generator (EDG) and Unit 3 Train A EDG automatically started and energized their respective 4.16 kV Class 1E Buses. 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 emergency AC electrical power systems. All systems operated as expected. Per the Emergency Plan, no classification was required due to the event. Units 1 and 3 both remain in Mode 1 at 100 percent power. Unit 2 is currently in a refueling outage and defueled. The 4.16 kV Class 1E Buses in Unit 2 were not affected by the de-energization of the start-up transformer since it was not aligned as normal power for Unit 2. The cause of the start-up transformer being de-energized is under investigation. The NRC Resident Inspectors have been informed. * * UPDATE ON 12/3/21 AT 1652 EST FROM MATT BRADFIELD TO KERBY SCALES * * * As a result of the Loss of Power on the Unit 1 Train B 4.16 kV Class 1E Bus, the B Auxiliary Feedwater Pump automatically started, as expected. The B Auxiliary Feedwater Pump was not needed for steam generator water level control and no auxiliary feedwater valves repositioned. The B Auxiliary Feedwater Pump did not supply feedwater to the steam generators. The NRC Resident Inspector will be notified. Notified R4DO (Taylor).
ENS 55531	20 August 2021 16:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 - Pressure Regulator Valve Failure	The following is a summary of the report provided by Engine Systems Inc.: The vendor supplied information to the NRC involving a defect and/or failure related to a pressure regulator valve installed on an emergency diesel generator at Brunswick Nuclear Power Plant. The valve did not properly regulate starting air pressure and allowed equalization of inlet pressure to outlet. Subsequent investigation by the vendor revealed a raised edge on the metal seating surface of the valve that caused the PTFE (Teflon) seat to tear. Equalization of starting air pressure is undesirable since it may inhibit operation of the downstream starting air solenoid valve, thus compromising the ability of the emergency diesel generator to start and support safety-related loads. The evaluation is complete. This Part 21 applies only to valves in the Brunswick Nuclear Power Plant. Corrective Actions: Brunswick continues to monitor the outlet pressure from the regulator and verify the inlet and outlet pressures have not equalized. The vendor also recommends that existing regulators have an inspection performed on-site and at the vendor. The vendor will add inspections to the dedication package for new and refurbished pressure regulator valves to verify a smooth, rounded transition at the valve seat of the throttling sleeve. Technical questions concerning this notification can be directed to Dan Roberts, Quality Manager, and John Kriesel, Engineering Manager.
ENS 55529	18 October 2021 00:30:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Emergency Diesel Generators Simultaneously Inoperable	At 1930 CDT on 10/17/2021, it was discovered both of the Unit 2 Emergency Diesel Generators were simultaneously INOPERABLE with a requirement to have one OPERABLE train; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10CFR 50.72(b)(3)(v). Offsite power was OPERABLE during this event. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. The cause was corrected and both Emergency Diesel Generators are currently operable. * * RETRACTION ON 12/13/2021 AT 1607 EST FROM CARLOS PARADA TO LLOYD DESOTELL * * * The following information was provided by the licensee via email: This is a retraction of Event Notification EN55529 in accordance with 10 CFR 50.72(b)(3)(v)(D) made by the Prairie Island Nuclear Generating Plant on October 18, 2021. The original notification stemmed from a loss of power to the non-safety related Unit 2 Emergency Diesel Generator (EDG) starting air compressors. The resulting pressure decay in the EDG starting air receivers led to a decision to declare both EDGs inoperable. A subsequent engineering evaluation has provided reasonable assurance that the Unit 2 EDGs were operable and capable of performing their safety function during the time power was lost. The NRC Resident Inspector has been notified. The HOO notified R3DO (Skokowski).
ENS 55490	24 September 2021 05:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	General Electric Hitachi Trip Units Rivets Popping Off	Based upon a 10 CFR 21.21(b) transfer notification from General Electric Hitachi (GEH), Southern Nuclear Operating Company's (SNC) Joseph M. Farley Nuclear Plant (FNP) has determined there is evidence a Substantial Safety Hazard could have been created by the failure of the rivets installed on certain EC Trip Units if they were left uncorrected. These EC Trip Units are a subcomponent of all five (5) emergency diesel generator control panel supply breakers at FNP. This defect was identified and the components were repaired by GEH before being installed in the plant. These defective EC Trip Units never posed a challenge to the safe operation of FNP. The NRC Senior Resident Inspector at FNP has been notified.
ENS 55459	13 September 2021 22:22:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition of Fire Safe Shutdown Equipment	On September 13, 2021, at 1822 EDT, an apparent non-compliance with 10 CFR 50, Appendix R, section III.G.2 (separation of redundant fire safe shutdown equipment) was identified. Specifically, it was determined that some Emergency Diesel Generator (EDG) cables may be susceptible to a hot short/spurious operation to the close circuit. A spurious closure of the emergency bus normal supply breakers after the EDG is powering the bus could result in non-synchronous paralleling, EDG overloading, or EDG output breaker tripping due to faulted power cable from normal supply breaker. The spurious closure of the normal supply breakers is not currently addressed in the Appendix R Report or previous Multiple Spurious Operations (MSO) analysis. This condition is associated with the Appendix R safe-shutdown function of the Emergency Power System. The Emergency Power System is considered operable but not fully qualified for its safety-related design function. The following fire areas are impacted: 1) Fire Area 13, Unit 1 Normal Switchgear Room 2) Fire Area 46, Unit 1 Cable Tray Room 3) Fire Area 3, Unit 1 Emergency Switchgear and Relay Room 4) Fire Area 2, Unit 2 Cable Vault and Tunnel Until this condition is analyzed, Surry has implemented mitigating actions in the above fire areas. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60-day written report pursuant to 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as CR (condition report) 1180502. The NRC Resident Inspector has been notified of this event. Mitigating actions include posting fire watches in the affected areas.
ENS 55436	29 August 2021 23:04:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Safety System Actuation	At 1804 CDT on 8/29/2021, Waterford 3 Steam Electric Station (WF3) experienced a Loss of Off Site Power event due to Hurricane Ida (See EN #55435). This event caused an automatic actuation of Emergency Diesel Generators Trains A and B. Both Emergency Diesel Generators started as designed and both are currently operating normally supplying power to their respective Class 1E Safety Busses. This automatic actuation is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). Prior to the loss of offsite power, WF3 was in progress of performing a plant cooldown in accordance with procedural guidance. As part of this cooldown and after entering Mode 4, all Safety Injection Tanks were isolated. As a result of losing offsite power, Reactor Coolant System Temperature increased above 350F which is above the temperature requirements for Mode 4. Safety Injection Tanks are required to be unisolated and OPERABLE in Mode 3. Therefore, with no Safety Injection Tanks OPERABLE, this constituted an event or condition that could have prevented the fulfillment of a safety function and the unit entered Technical Specification 3.0.3. The unit was in Technical Specification 3.0.3 for approximately 43 minutes from 1805 CDT until 1848 CDT when Mode 4 conditions were re-established. This event or condition that could have prevented the fulfillment of a Safety Function is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D). While continuing to perform the Reactor Coolant System Cooldown and prior to placing Shutdown Cooling Train in service, it became necessary to start one train of Emergency Feedwater. Emergency Feedwater Train A was manually started at 1847 CDT to feed the Steam Generators and was secured at 1947 CDT. Emergency Feedwater Train A started and operated normally during this period. This manual actuation is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector. * * UPDATE AT 1549 EDT ON OCTOBER 25, 2021 FROM CHANTEL HATTAWAY TO BRIAN P. SMITH * * * The purpose of this notification is to revise Event Notification Report (EN) 55436 to include a partial retraction. On August 29, 2021, Waterford Steam Electric Station, Unit 3 (WF3) experienced a loss of offsite power (LOOP) event due to Hurricane Ida. Prior to the LOOP, WF3 had shutdown to Mode 3 (Hot Standby) in anticipation of the LOOP and was performing a plant cooldown in accordance with procedural guidance. When Mode 4 (Hot Shutdown) was achieved, all Safety Injection Tanks (SITs) were isolated as part of the plant cooldown. After the LOOP, Reactor Coolant System (RCS) temperature increased and the Core Exit Thermocouples (CETs) indicated that RCS temperature had exceeded 350 degrees F. Based on the CETs, this was above the temperature requirements for Mode 4 and, as such, WF3 declared entry into Mode 3. The SITs are required to be unisolated and Operable in Mode 3. Since no SITs were Operable at that time, it was determined that this constituted an event or condition that could have prevented the fulfillment of a safety function and included this as part of the EN 55436 report in accordance with 10 CFR 50.72(b)(3)(v)(D). An engineering evaluation has subsequently been performed to validate whether the RCS temperature excursion following the LOOP actually reached 350 degrees F. As defined in WF3 Technical Specification (TS) Table 1.2, Operational Mode temperatures are a function of RCS average temperature (Tavg), not just the indicated temperature of the CETs. Based on the calculated Tavg using validated temperatures, it was concluded that 350 degrees F was not reached. Thus, WF3 remained in Mode 4 following the LOOP and there was no event or condition that could have prevented the fulfillment of a safety function that was reportable pursuant to 10 CFR 50.72(b)(3)(v)(D). The remainder of EN 55436 remains correct and unchanged. The licensee notified the NRC Resident Inspector. Notified R4DO (Pick)
ENS 55435	29 August 2021 23:12:00	10 CFR 50.72(a)(1)(i), Emergency Class Declaration	Notification of Unusual Event Due to Loss of Offsite Power	Waterford 3 shut down the reactor in preparation for Hurricane Ida landfall prior to this event. At 1812 CDT, Waterford 3 declared a notification of unusual event under EAL S.U. 1.1 due to a loss of offsite power as a result of hurricane Ida. Plant power is being provided via emergency diesel generators. The NRC Activated at 2016 EDT with Region IV in the lead. Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, CISA Central, USDA Ops Center, EPA Emergency Ops Center, DHS Nuclear SSA (email), FEMA NWC (email), and FEMA NRCC SASC (email). * * UPDATE ON 9/1/21 AT 0132 EDT FROM ALEX SANDOVAL TO BRIAN P. SMITH * * * At 2345 CDT on 8/31/21, Waterford 3 terminated their notification of unusual event under EAL S.U. 1.1. Offsite power has been restored to both safety-related electrical buses. The NRC remains Activated with Region IV in the lead while reviewing additional criteria to exit Activation. Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, CISA Central, USDA Ops Center, EPA Emergency Ops Center, DHS Nuclear SSA (email), FEMA NWC (email), and FEMA NRCC SASC (email), R4DO (Josey), IR MOC (Kennedy), NRR EO (Miller), R4 (Lantz).
ENS 55423	22 August 2021 09:29:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	HPCI Declared Inoperable	At 0529 EDT on August 22, 2021, HPCI ((High Pressure Coolant Injection System)) was declared inoperable due to receiving the HPCI Inverter Circuit Failure annunciator. The cause of the annunciator was a fuse failure. The cause of the fuse failure is unknown at this time and is under investigation. Concurrent with the HPCI fuse failure was a similar fuse failure within the Division 2 EDG ((emergency diesel generators)) Load Sequencer which renders the Division 2 EDGs inoperable. Relation to the HPCI issue is unknown and is part of the investigation. The RCIC ((Reactor Core Isolation Cooling System)) was verified operable per Tech Spec 3.5.1 E.1. In addition, offsite circuits were verified operable per Tech Spec 3.8.1.B. Division 1 EDGs remain operable. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. There was no impact on the health and safety of the public or plant personnel. The Senior NRC Resident Inspector has been notified.
ENS 55375	22 July 2021 21:51:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition of Fire Safe Shutdown Equipment	On July 20, 2021, at 1707 EDT, an apparent non-compliance with 10 CFR 50, Appendix R, section III.G.2 (separation of redundant fire safe shutdown equipment) was identified. This issue was initially categorized as not affecting train separation or the ability of the equipment to perform their Design Basis functions. The original concern was entered into the licensee's Corrective Action Program as CR1177199. Subsequently, on July 22, 2021, at 1751 EDT, a further review of the affected control circuits for the Unit 1 and Unit 2 Emergency Diesel Generator (EDG) output breakers and emergency bus feeder breakers identified a concern that breaker position interlocks routed to or through non-safety related components or spaces may affect the ability to provide emergency power on the affected unit due to impacts on the control power circuits during an Appendix R fire associated with a loss of offsite power. The following are the affected fire areas: - Unit 1 and Unit 2 Turbine Buildings - Unit 1 and Unit 2 Cable Spreading Rooms - Unit 1 and Unit 2 Normal (307) Switchgear Rooms This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60-day written report pursuant to 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as CR 1177399. The NRC Resident Inspector has been notified of this event.
ENS 55347	11 May 2021 16:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 Initial Report - Deviation Identified Loose Zinc Anodes Installed in Heat Exchangers for Diesel Generators	The following is a summary of the defect described in an initial report received from the vendor via facsimile: The vendor notified the NRC of a defect involving two instances of loose or dislodged zinc anodes (P/N 1335BEM2P) supplied by the vendor. The anodes are installed in the cooling water enter and exit and return channels of the jacket water heat exchanger of the emergency diesel generator. Each heat exchanger contains eight zinc anode assemblies that consist of a zinc rod threaded into a steel pipe plug. The zinc acts as a sacrificial anode to protect the pressure boundary metals from degradation due to galvanic corrosion. In the case of a loose or dislodged zinc anode, the zinc rod may become foreign material that remains trapped in the vessel, potentially impacting and damaging the tube ends where they project from the tube sheet. Alternatively, an anode located in the exit channel may by carried away from the heat exchanger and potentially damage downstream components. The corrective action being recommended are as follows: It is recommended to perform an inspection to verify tightness at the zinc rod to pipe plug interface. The rod should be threaded into the pipe plug snug tight and, if required, may be tightened to a maximum of 15 ft-lbs. If desired, the zinc rod may be removed and eliminated from the assembly leaving only the steel pipe plug. Inclusion of the zinc anode is not required for the EDG heat exchanger application at TVA-Browns Ferry. This component was supplied to the Browns Ferry Nuclear Power Plant. Point of contact: Dan Roberts Quality Manager Engine Systems Inc. 175 Freight Rd. Rocky Mount, NC 27804 (252) 977-2720
ENS 55306	15 June 2021 17:30:00	10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release	Offsite Notification	At 1230 CDT a report was made to the State of Nebraska Department of Environment and Energy (NDEE) based on the analytical report for soil samples from the area surrounding the removed FO-1, Emergency Diesel Generator Fuel Oil Storage Tank, and the removed FO-32, TSC/Security Fuel Oil Tank. The tanks were removed as part of Fort Calhoun Station decommissioning and soil samples were tested due to soil discoloration at the time the tanks were pulled. The soil contamination levels are from the historic use of the tank. The contamination levels are above the lab reporting limits and thereby reportable to the State of Nebraska Department of Environment and Energy. The NDEE will determine what, if any, remediation may be required. The state NDEE requested the District utilize their Spill Form because this is the simplest method of State notification for tanks exempted due to 40CFR280.10(c)(4). No active petroleum spills are in progress and appropriate remediation actions will be taken in accordance with Nebraska State regulation and guidance. The licensee notified the NRC Region IV Office.
ENS 55265	19 May 2021 10:15:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Automatic Reactor Trip Due to High Pressurizer Pressure	At 0315 MST on May 19, 2021, Unit 2 reactor automatically tripped during testing of the Plant Protection System. The Reactor Protection System actuated to trip the reactor on High Pressurizer Pressure, although no plant protection setpoints were exceeded. Main Steam Isolation Signal (MSIS), Safety Injection Actuation Signal (SIAS), and Containment Isolation Actuation Signal (CIAS) were received. No injection of water into the Reactor Coolant System occurred. Auxiliary Feedwater Actuation Signals (AFAS) 1 and 2 actuated on low Steam Generator water level post trip as designed. This event is being reported as a reactor protection system and a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Following the reactor trip, all (Control Element Assemblies) CEAs inserted fully into the core. All systems operated as expected. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. Unit 2 is stable and in Mode 3. Steam Generator heat removal is via the class 1 E powered motor driven auxiliary feedwater pump and Atmospheric Dump Valves. The NRC Senior Resident Inspector has been informed. * * UPDATE ON 5/19/21 AT 1351 EDT FROM JASON HILL TO BRIAN P. SMITH * * * The Unit 2 reactor tripped because of actual High Pressurizer Pressure that occurred as a result of a Main Steam Isolation Signal actuation. At 0337 MST, both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) were made inoperable when the injection valves were overridden and closed in accordance with station procedures. At 0346 MST, in accordance with station procedures, both trains of Containment Spray, LPSI, and HPSI pumps were overridden and stopped, rendering Containment Spray inoperable as well. This represents a condition that would have prevented the fulfillment of a safety function required to mitigate the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). Additionally, at the time of the Safety Injection Actuation Signal (0315 MST), both trains of Emergency Diesel Generators actuated as required and both 4160 VAC busses remained energized from off-site power. The NRC Senior Resident Inspector has been informed. Notified R4DO (Young) * * UPDATE ON 7/02/21 AT 1943 EDT FROM YOLANDA GOOD TO JEFFREY WHITED * * * The inoperability of both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) and both trains of Containment Spray (CS) following the Unit 2 reactor trip has been determined to be an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Additionally, inoperability of both trains of HPSI resulted in a reportable condition that could prevent fulfillment of its credited safety function to maintain the reactor in a safe shutdown condition per 10 CFR 50. 72(b)(3)(v)(A). The additional reporting criteria were discovered during review of the event and corresponding safety analyses. The NRC Senior Resident Inspector has been informed. Notified R4DO (Werner)

Semantic search

Query log

Navigation menu