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	Site	Start date	Title	Description
05000423/LER-2019-001	Millstone	13 February 2020	LER 2019-001-00 for Millstone Power Station Unit 3 Regarding Plant Shutdown Required by Technical Specifications, Emergency Diesel Generator Exceeded Allowed Outage Time
05000289/LER-2017-004	Three Mile Island Three Mile Island Unit 1	5 February 2018	Items Nonconforming to Design for Tornado Missile Protection LER 17-004-00 for Three Mile Island, Unit 1, Regarding Items Nonconforming to Design for Tornado Missile Protection	On December 6, 2017 Three Mile Island Unit 1, determined that both Emergency Diesel Generators do not conform with the licensing bases for protection against tornado generated missiles. The vent on the common Fuel Oil Supply Tank that serves both Emergency Diesel Generators could be damaged by debris generated from a tornado that could affect emergency diesel generator operation. An extent of condition review identified three additional items that are nonconforming to the design for tornado missile protection: vent stacks for each Emergency Diesel Generator Day Fuel Tanks, the Borated Water Storage Tank and steam piping near the main feed pumps that could affect Secondary Pressure Control. Upon determination of the initial nonconformance for the Fuel Oil Supply Tank Vent on December 6, 2017, both Emergency Diesel Generators were declared inoperable. Compensatory measures were put and verified in place in accordance with the NRC Enforcement Guidance Memorandum EGM 15-002, both emergency diesel generators were returned to an operable but nonconforming status and an 8 hour ENS Notification was made to the NRC. This condition has been in existence since original licensing of the plant. It is not known if it was overlooked or considered acceptable at the time of the original licensing process. There are no actual consequences as a result of the nonconforming conditions.
05000391/LER-2017-005	Watts Bar Watts Bar Nuclear Plant. Unit 2	25 January 2018	Unplanned Emergency Core Cooling System Injection into the Reactor Coolant System due to Personnel Error LER 17-005-00 for Watts Bar, Unit 2, Regarding Unplanned Emergency Core Cooling System Injection into the Reactor Coolant System due to Personnel Error	On November 26, 2017. at 1225 Eastern Standard Time (EST), the Watts Bar Nuclear Plant (WBN) Unit 2 experienced an unplanned Emergency Core Cooling System (ECCS) discharge to the Unit 2 Reactor Coolant System (RCS) while de-pressurized. in Mode 5. with the Pressurizer vented to the Pressurizer Relief Tank. ECCS injection via the Boron Injection flow path occurred during planned Safety Injection system Engineered Safety Features Actuation System (ESFAS) testing. The Boron Injection flow path should have been isolated and should not have resulted in any injection flow to the Unit 2 RCS. The condition was promptly corrected by operator actions based on observed plant conditions. The cause of this event is that an Operator improperly used a Caution Order to determine the configuration of the breaker for the Boron Injection Tank outlet valve. Correct Component Verification was not utilized as required. and the current position of the breaker in the field was not validated to support testing. Corrective actions for this event include revising procedures to ensure the breakers associated with the boron injection flow path will be tagged open during ESFAS testing and that lessons learned related to this event are communicated to operating crews. An evaluation on the use of Caution Orders for off normal equipment positions will be performed . NRC FORM 330604-2O'
05000219/LER-2017-005	Oyster Creek	9 October 2017 3 January 2018	1 OF 3 LER 17-005-00 for Oyster Creek Nuclear Generating Station Regarding Failure of the Emergency Diesel Generator #2 During Surveillance Testing Due to a Broken Electrical Connector	On 10/09/17, during the bi-weekly Emergency Diesel Generator (EDG) #2 Load Test, a Generator Lockout signal (86G) was received which tripped the EDG output breaker. This failure resulted in EDG #2 being declared inoperable, and entry into an unplanned 7-day Limiting Condition for Operation (LCO) at 0312. Troubleshooting identified a broken electrical ring lug connector on a current transformer that provides an input to the protective relay logic. The investigation determined the connector failure was due to fatigue cracking that was initiated by stresses caused by bending and twisting of the electrical lug beyond limits specified in industry guidelines. The ring lug was most likely distressed during initial installation in the 1990's. The condition that led to the EDG #2 trip existed for greater than the allowed outage time in the plant's Technical Specification (TS) and is reportable under 10 CFR 50.73(a)(2)(i)(B).
05000296/LER-2017-002	Browns Ferry	29 December 2017	4kV Shutdown Board Potential Transformer Primary Fuses Do Not Coordinate with Secondary Fuses LER 17-002-00 for Browns Ferry Nuclear Plant, Unit 3 Regarding 4kV Shutdown Board Potential Transformer Primary Fuses Do Not Coordinate with Secondary Fuses	On November 1, 2017, at approximately 1425 Central Daylight Time, during an extent of condition review, a 4kV Shutdown Boards (SD BD) do not selectively coordinate with upstream 0.5 amp primary fuses for fault currents greater than 30 amps on the 120 V secondary. Cable fire damage could cause an affected SD BD to spuriously disconnect from off-site power, and could cause a spurious, maintained under-voltage trip signal. The under-voltage trip signal would prevent motor load operation on the board whether on off-site 4kV SD BDs could be affected. In fire area 21, 4kV SD BDs 3EA and 3EB could be affected. This condition was determined to be a legacy issue dating to the original design of the plant. The most likely cause is lack of rigorous oversight of the vendor during the preparation and subsequent issuance of the fuse evaluation for the four Unit 3 4kV SD BDs. The required coordination studies have since been performed and a vendor oversight process has been added to TVA procedures. Compensatory measures (hourly fire watches) have been put in place for affected fire areas. Additional corrective actions include issuing an Engineering Change Package to replace the Unit 3 4kV SD BDs primary 0.5 amp PT fuses with 1 amp fuses of the same type.
05000390/LER-2017-016	Watts Bar Nuclear Plant. Unit 1	20 December 2017	System Actuations Due to Opening of Feeder Breaker to the 1 B-B 6.9 kV Shutdown Board	On December 20. 2017, at 1040 Eastern Standard Time (EST), the Watts Bar Nuclear Plant (WBN) 1B-B 6.9kV Shutdown Board (SDBD) normal feeder breaker opened. The loss of voltage to the 1B-B SDBD resulted in the start of the 1 B-B Motor Driven Auxiliary Feedwater (MDAFW) pump. the Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) pump. and the start of all four Emergency Diesel Generators (EDGs). Power was restored to the 1B-B SDBD when it loaded on to its associated EDG. Following initial investigation, the 1B-B SDBD was transferred to its alternate offsite power source at 1217 EST. At 1230 EST. the 1 B-B SDBD alternate feeder breaker opened, with a plant response that was similar to the first loss of power. Restoration of normal offsite power to the 1B-B SDBD was completed at 1654 EST. This event is being reported as a safety system actuation and as an event or condition that could have prevented fulfillment of a safety function related to containment temperature being outside Technical Specification limits. Both loss of voltage events to the 1B-B SDBD were caused by poor contact of the B and C phases of the protective relay potential transformer drawer secondary connections which supplies the degraded and loss of voltage relays. The mounting block that houses the secondary pins was able to be trimmed, resulting in an improvement of the secondary connection. The mounting blocks for the secondary connections on SDBDs 1A-A, 2A-A, and 2B-B will be inspected and modified. if required, during future equipment outages. The procedure associated with inspection of the secondary connections will be revised. NRC FORM He :2-2:- APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 comments regarding burden estimate to the Information Services Branch (T-2 F43). U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov. and to the Desk Officer. Office of Information and Regulatory Affairs, used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
05000286/LER-2017-004	Indian Point	3 November 2017 20 December 2017	Reactor Trip Due to Main Generator Loss of Field LER 17-004-00 for Indian Point Unit 3, Regarding Reactor Trip Due to Main Generator Loss of Field	On November 3, 2017, at 2022 hours, with reactor power at 100 percent, Indian Point Unit 3 experienced an automatic reactor trip on a turbine trip, which was in response to a main generator trip. The main generator trip was initiated by actuation of the Generator Protection System due to a main generator loss of field. All control rods fully inserted and all required safety systems functioned properly. The plant was stabilized in hot standby with decay heat being removed by the main condenser. The Auxiliary Feedwater System (AFWS) automatically started as expected on steam generator low level to provide feedwater flow to the steam generators. The plant was stabilized in hot standby with decay heat being removed by the main condenser. The direct cause of the loss of main generator field was a failed Thyristor Firing Module drawer which affected proper operation of the redundant Thyristor Firing Module drawer. The root cause was determined to be that the Automatic Voltage Regulator (AVR) Firing Module power supplies have a latent design vulnerability where shared common output nodes are not isolated after a failure. A plant modification is proposed that will eliminate the condition by electrically isolating the AVR Firing Module power supplies upon failure. This event had no effect on the public health and safety. The event was reported to the Nuclear Regulatory Commission (NRC) on November 3, 2017 under 10 CFR 50.72(b)(2)(iv)(B) and 50.72(b)(3)(iv)(A) as an event that resulted in the automatic actuation of the Reactor Protection System when the reactor is critical and a valid actuation of the AFWS.
05000416/LER-2017-007	Grand Gulf	12 December 2017	Engineered Safety Feature System Actuations due to the loss 01 Engineered Safety Features Transformer 11	At approximately 0918 hours on Tuesday, December 12, 2017, while operating in MODE 1 at approximately 18 percent power, the Grand Gulf Nuclear Station (GGNS) experienced a loss of the Engineered Safety Features (ESF) Transformer 11 which was powering the Division 1 ESF bus. Subsequently, the station experienced an automatic start of the Division 1 Emergency Diesel Generator and the partial isolation of the primary and secondary containment buildings. Both of these events were expectedand as designed. The direct cause of ESF actuations was the loss of ESF Transformer 11. The cause of the transformer loss is under investigation at this time and this licensee event , report will supplemented upon completion of GGNS's causal analysis. Additionally, GGNS experiented an unrelated isolation of the Reactor Core Isolation Cooling System upon restoration of power. The isolation of the Reactor Core Isolatigh Cooling System did not result in a loss of safety function. The cause of this isolation is under investigation and will be documented in accordance with the.GGNS corrective action program. This event is reportable to the NRC in accordanCe with 10 CFR 50.72(b)(3)(iv) and 10 CFR 50.73(a)(2)(iv)(A) as an event or condition resulting in a valid actuation of a ESF system. Grand Gulf Nuclear Station, Unit 1 05000 416 . NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/3112020 (4-2017) Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 so RkG,„ LICENSEE EVENT REPORT (LER) r F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to y n4 CONTINUATION SHEET Infocollects.Resource@nrc.gov: and to the Desk Officer: Office of Information and .i: Regulatory Affairs, NEOB-10202. (3150-0104). Office of Management and Budoet, Washington, DC 20503: If a means used to impose an information collection does not c's, T (See NUREG-1022, R.3 for instruction and guidance for completing this form display a currently valid OMB control number, the NRC may not conduct or sponsor. and a N*,......0, htto://vAmnrc.00virP-adiriq-rmidoc-collectionsinureosistaff/sr1022/r3A . person is not required to respond to, the information collection. Contents 1 DESCRIPTION 2 REPORTABI LITY 3 CAUSE 4 NRC FORM 5 CORRECTIVE ACTIONS 6 SAFETY SIGNIFICANCE 7 PREVIOUSLY SIMILAR EVENTS DESCRIPTION At approximately 0918 hours on Tuesday, December 12, 2017, while operating in MODE 1 at approximately 18 percent power, the Grand Gulf Nuclear Station (GGNS) experienced a loss of the Engineered Safety Features (ESF) Transformer 11 (EB) which was powering the Division 1 ESF bus (EA): The transformer experienced an instantaneous ground resulting in a transformer lockout and loss of power to the ESF bus. Subsequently, the station experienced an automatic start of the Division 1 Emergency Diesel Generator (EK) and the partial isolation of the primary and secondary containment buildings. Both of the system actuations were expected responses to a loss of ESF bus and both systems responded as designed. The direct cause of ESF actuations was the loss of ESF Transformer 11. Additionally, GGNS experienced an unrelated isolation of the Reactor Core Isolation Cooling System (BN) upon restoration of power. The' isolation of the. Reactor Core Isolation Cooling System did not result in a loss of safety function. The cause of this isolation is under investigation and will be documented in accordance with the GGNS corrective action program. REPORTABI LITY This event is reportable to the NRC in accordance with 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.73(a)(2)(iv)(A) as an event or condition resulting in a valid actuation of a ESF system. The 10 CFR 50.72 reporting requirements were met with the completion of Emergency Notification System (ENS) Notificatibn 53115, at 1740 hpurs eastern standard time on December 12, 2017. CAUSE Direct Cause: The direct cause of the ESF actuation was the loss of ESF Transformer 11 and the opening of the transformer feeder breaker due to an instantaneous ground. Apparent Cause: The most probable cause is a ground on one of the feeder cables to ESF Transformer 11. However, the investigation and causal analysis is ongoing at this time and this licensee event report will be supplemented upon completion of the GGNS causal analysis. NRC FORM (6-2016) 366A U.S. NUCLEAR. REGULATORY COMMISSION LICENSEE. EVENT REPORT (LER) CONTINUATION 'SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form htto://www.nrc.coWreadino-rm/doc-collectionsinureos/staff/sr1022/r3/) APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/3112020 Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington. DC 20555-0001, or by e-mail to Infoccillects.Resource@nrc.gov, and to the Desk Officer. Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington. DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number. the NRC may not conduct or sponsor. and a person is not required to respond to, the information collection. 2. DOCKET 3. LER NUMBER 05000.416 CORRECTIVE ACTIONS Spare Essential Transformer 21 was placed into service and normal power was restored. The investigation and causal analysis is ongoing and this licensee event report will be supplemented upon completion of GGNS's causal analysis. The planned corrective actions will be included in the corrective action program and may be changed in accordance with the program. .": SAFETY SIGNIFICANCE There were no nuclear safety consequences or radiological consequences as a result of this event. No Technical Specification Safety Limits were violated. Upon the loss of Engineered Safety Feature Transformer 11 all required accident mitigation ESF components responded as designed. The isolation of the Reactor Core Isolation Cooling System, although unexpected, did not adversely impact the plant's ability to respond to the event. PREVIOUSLY SIMILAR EVENTS Protective Relaying Circuitry on the "B" Main Transformer Transformer Wiring Entergy has reviewed the events listed in the licensee event reports (LER) documented above to determine if the corrective actions should have prevented the event documented in this LER. Based on a preliminary evaluation it has been concluded the established corrective actions would not have prevent this event. Entergy's investigation into the cause of this event and the development of corrective actions to preclude recurrence are ongoing. This section will be supplemented at the conclusion of this effort.
05000397/LER-2017-007	Columbia	3 October 2017 30 November 2017	Valve Closure Results in Momentary Increase in Secondary Containment Pressure LER 17-007-00 for Columbia Generating Station Regarding Valve Closure Results in Momentary Increase in Secondary Containment Pressure	On October 3, 2017 at 0800 PDT, Secondary Containment (Reactor Building) became inoperable due to pressure increasing above the Technical Specification (TS) limit of -0.25 inches of water gauge (inwg). While the plant was at 100% power, a Reactor Building exhaust valve unexpectedly closed, resulting in a loss of Secondary Containment for approximately two minutes. Secondary Containment was declared inoperable and TS Action Statement 3.6.4.1.A was entered. The Control Room operators reopened the Reactor Building exhaust valve and pressure returned to within limits automatically. Secondary Containment was declared operable at 0810 PDT and TS Action Statement 3.6.4.1.A was exited. The event was reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D) as Event Notification #52999. The apparent cause of the event is a surface degradation on the lower stab of an electrical disconnect causing a momentary high resistance when the cubicle door is opened. This event occurred during performance of thermography in the cubicle.
05000425/LER-2017-002	Vogtle	17 November 2017	Valid Undervottage Condition results in Automatic Actuation of a Safety System LER 17-002-00 for Vogtle, Unit 2, Regarding Valid Undervoltage Condition Results in Automatic Actuation of a Safety System	On September 26, 2017, at approximately 0543 EDT, while performing 2B Emergency Diesel Generator (EDG) and Emergency Safety Function Actuation System (ESFAS) testing, a valid undervoltage actuation signal was sent to the Unit 2 B-Train Emergency Diesel Generator. The 2B AC emergency bus (2BA03) was load shed, the 2B EDG automatically started, and tied to 2BA03 The 2BA03 bus was loaded by the automatic load sequencer. The actuation was identified by the Control Room operators and the 2B EDG was locally monitored while in service. An air leak on the shutdown logic board caused a non-emergency trip to actuate when the non-emergency trips were unblocked. The reactor was in Mode 6 at the time of the event and not challenged throughout the event. Decay heat removal and spent fuel pool cooling were not challenged throughout the event. Because this event resulted in the automatic actuation of a system listed in 10 CFR 50.73(a)(2)(iv)(B), this event is reportable under 10 CFR 50.73(a)(2)(iv)(A). This event had no adverse effect on the health and safety of the public, and is of very low safety significance. Unit 1 was unaffected.
05000293/LER-2017-012	Pilgrim	13 September 2017 13 November 2017	Start-Up Transformer Degraded Voltage Relay Found Outside Technical Specification Limit LER 17-012-00 for Pilgrim Nuclear Power Station Regarding Start-Up Transformer Degraded Voltage Relay Found Outside Technical Specification Limit	On September 13, 2017 with the unit at 100 percent power, Pilgrim Nuclear Power Station discovered during calibration of four Start-Up Transformer Degraded Voltage Relays (ABB model number ITE- 27N) after the relays had been removed from the plant, relay 127A-604/1 had an as-found setpoint value outside the Technical Specifications Table 3.2.B limit of 110.6 plus or minus 0.56 Volts ac (Vac) (110.04 - 111.16); the as-found value was 109.90 Vac. This relay was replaced with a spare relay that was calibrated prior to its installation. Pilgrim Nuclear Power Station is submitting this Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(i)(B) - Any operation or condition that was prohibited by the plant's Technical Specifications. This event was not risk significant. There was no threat to public health and safety from this condition.
05000395/LER-2017-004	Summer	9 November 2017	ACTUATION OF 'A' EMERGENCY DIESEL GENERATOR LER 17-004-00 for V.C. Summer, Unit 1, Regarding Actuation of 'A' Emergency Diesel Generator	On September 11, 2017 at 1648. the VCSNS Unit I 'A' emergency diesel generator (EDG) was actuated. VCSNS Unit I was and continued to operate in Mode 1 at 100% reactor power. The EDG actuation was caused by a storm induced perturbation on the off-site power system. The duration of the fault was longer than it should have been due to a malfunction of a transmission system relay. The perturbation cleared and off-site voltage was returned to normal within the designed recovery time limit. The bus continued to be carried by the off-site source and the EDG output breaker remained open. The station review of this event has shown that plant response was as designed and that no safety consequences occurred. NRC FORM 365 (04-2017)
05000341/LER-2017-005	Fermi	3 November 2017	Non-Functional Mechanical Draft Cooling Tower Fan Brakes Leads to HPCI Being Declared Inoperable and Loss of Safety Function LER 17-005-00 for Fermi 2 Regarding Non-Functional Mechanical Draft Cooling Tower Fan Brakes Leads to HPCI Being Declared Inoperable and Loss of Safety Function	At 1000 EDT on September 9, 2017, the Division 2 Mechanical Draft Cooling Tower (MDCT) fans were declared inoperable due to loss of output from the over speed fan brake inverter. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS) and the Emergency Equipment Cooling Water (EECW) system. The Division 2 EECW system cools the High Pressure Coolant Injection (HPCI) system room cooler. As a result, the non-functionality of the fan brakes lead to an unplanned HPCI inoperability. Since HPCI is a single train system designed to mitigate the consequences of a loss of coolant accident (LOCA), this event could have prevented the fulfillment of a safety function. The cause of the event was the failure of the Division 2 fan brake inverter. Corrective Actions were taken to replace the inverter and returning the MDCT fans, the UHS, EECW and HPCI to service on September 9, 2017 at 2351 EDT. A failure modes evaluation was performed by the vendor with no direct cause of the failed output determined. The fan brake system is only required for a design basis tornado and there was no credible tornado threat during this event. The HPCI system is not required to mitigate a design basis tornado. The safety significance of this event is very low and there were no radiological releases associated with this event.
05000390/LER-2017-012	Watts Bar	23 October 2017	Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications LER 17-012-00 for Watts Bar, Unit 1, Regarding Error in Plant Emergency Procedures Leads to a Condition Prohibited by the Technical Specifications	On August 23. 2017. Watts Bar Nuclear Plant (WBN) identified that procedures 1-E-1 and 2-E-1, Loss of Reactor or Secondary Coolant, contained steps to manually open 1-FCV-67-458 in the event of a Train A or B power failure. Opening 1-FCV-67-458 would result in the crosstie of Essential Raw Cooling Water (ERCW) Headers 2A and 1B, which would lead to providing flow to equipment not operating due to the loss of a train of power. On October 6. 2017. it was determined that for certain time periods, if a design basis accident had occurred on Unit 2 with a loss of offsite power concurrent with a train failure and with 1-FCV-67-458 opened, inadequate ERCW flow would have been available to remove decay heat after transfer to cold leg recirculation. This condition only affected operability of ERCW Train A. This is reportable as a condition prohibited by Technical Specification 3.7.8. The issue associated with this incorrect procedural step to cross-tie the ERCW trains in 1-E-1 and 2-E-1 was addressed as part of actions to resolve an ERCW design and procedure issue documented in Licensee Event Report (LER) 390-2017-009. This report, while related, identifies an issue that was not addressed in the prior LER. The cause was determined to be the incorrect application of a cross tie requirement associated with 10 CFR 50 Appendix R. Corrective action will be to include engineering in the review of procedures affected by complex design changes. NRC FORM 355 ;:;4-217' APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. or by e-mail to NEOB-10202. (3150-0104), Office of Management and Budget. Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number. the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
05000483/LER-2017-002	Callaway	15 August 2017 13 October 2017	Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design LER 17-002-00 for Callaway Plant, Unit 1, Regarding Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design	On August 15, 2017, Callaway Plant was in Mode 1 at 100 percent power. During evaluation of protection for safety-related equipment from the damaging effects of tornados, Callaway Plant personnel determined that the minimum-flow recirculation lines for the turbine-driven auxiliary feedwater pump (TDAFP) and both motor-driven auxiliary feedwater pumps (MDAFPs) could be damaged if a postulated tornado-generated missile were to penetrate the condensate storage tank (CST) valve house and strike the lines. In response, Operations declared all three auxiliary feedwater pumps inoperable. Compensatory measures were implemented consistent with Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance." Upon completion of the initial compensatory measures, the TDAFP and MDAFPs were declared Operable but nonconforming. Subsequent to the condition identified on August 15, 2017, continued investigation of tornado missile vulnerabilities led to discovery that the exposed steam exhaust stacks for the main steam safety valves and atmospheric steam dump valves, as well as the exposed vents for the diesel generator fuel oil storage and day tanks, were also susceptible to tornado missile damage to the extent that compliance with General Design Criterion 2 is not ensured. Compensatory measures were then promptly implemented for these conditions, as well, in accordance with EGM 15-002 such that the affected systems have been evaluated to be nonconforming but Operable. It has been determined that the identified noncomformances are an original plant design legacy issue. Long-term resolution for establishing compliance is under development and will be completed within the time frame described in the EGM.
05000341/LER-2017-004	Fermi	10 August 2017 9 October 2017	Inadequate Procedural Cuidance for Residual I feat Removal Complex Ventilation Systems Leads to Condition Prohibited by Technical Specifications and Loss if Safety Function LER 17-004-00 for Fermi 2 Regarding Inadequate Procedural Guidance for Residual Heat Removal Complex Ventilation Systems Leads to Condition Prohibited by Technical Specifications and Loss of Safety Function	On August 10, 2017, it was determined that inadequate procedural guidance for determining operability for ventilation support systems was being utilized. The Residual Heat Removal (RHR) switchgear and pump rooms have ventilation systems to maintain operability of the equipment in the rooms. Fermi 2 procedures had directed personnel to declare the supported equipment in the rooms inoperable due to nonfunclionality of the ventilation systems only if the room temperature exceeded the operability limit. Following discovery, a review of the RHR switchgear and pump room ventilation systems for the past three years was performed. The review identified multiple instances where the ventilation systems were nonfunctional and should have resulted in entry into applicable Technical Specifications (TS). Many of these instances resulted in operations or conditions prohibited by TS, since TS Required Actions were not completed within the Completion Times for restoration of affected equipment and plant shutdown. In addition, one instance was identified where the plant configuration was such that it could have prevented the fulfillment of the safety function to remove residual heat following a design basis accident. An engineering evaluation of this specific instance was performed and verified that the plant remained within its analyzed design basis. All other instances maintained one fully operable division of heat removal equipment such that no loss of safety function existed. There were no radiological releases associated with this event. The safety significance was determined to he very low. The cause of the event was inadequate procedural guidance. Immediate actions were taken to provide interim guidance related to the procedure. Corrective actions to revise the affected procedure have been completed.
05000263/LER-2017-005	Monticello	23 July 2017 20 September 2017	Diesel Generator Emergency Service Water System Automatic Transfer to Alternate Shutdown Panel LER 17-005-00 for Monticello Nuclear Generating Plant Regarding Diesel Generator Emergency Service Water System Automatic Transfer to Alternate Shutdown Panel	On July 23, 2017, the Emergency Diesel Generator (EDG) lube oil cooler immersion heater temperature switch failed (in the on position) which together with residual heat from 12 EDG operation actuated the EDG engine temperature switch resulting in a valid start signal to the 12 EDG Emergency Service Water (ESW) System pump and transfer of pump control to the Alternate Shutdown System panel. In accordance with 10 CFR 50.73(a)(2)(iv)(A) an automatic actuation of the ESW System is reportable. The apparent cause of the 12 EDG hot engine condition was a failed temperature switch for the immersion heater in concert with operating the 12 EDG. From the time of the problem identification through the 12 EDG surveillance test, lube oil temperature was monitored by Operations and never exceeded the normal operating band when the EDG was in the shutdown ready-to-start configuration and during the EDG monthly surveillance run. Therefore, the 12 EDG remained operable. The immediate corrective action was replacement of the failed temperature switch. A long-term corrective action is to evaluate the immersion heater and associated control circuit to determine if the temperature switch design and preventative maintenance frequency are appropriate. Corrective action will be taken based on the results of this evaluation.
05000382/LER-2017-002	Waterford Waterford Steam Electric Station, Unit 3	17 July 2017 18 September 2017	Automatic Reactor Scram due to the Failure of Fast Bus Transfer Relays to Automatically Transfer Station Loads to Off- Site Power on a Main Generator Trip LER 17-002-00 for Waterford, Unit 3, Regarding Automatic Reactor Scram due to the Failure of Fast Bus Transfer Relays to Automatically Transfer Station Loads to Off-Site Power on a Main Generator Trip	On July 17, 2017, at 1606 CDT, Waterford 3 experienced an automatic reactor scram due to a loss of forced circulation, which was the result of a loss of off-site power to the safety and non-safety electrical busses. Prior to the scram, plant operators manually tripped the main turbine and generator due to overheating of the isophase bus duct due to the failure of a shunt assembly connection in the duct to Main Transformer 'B'. The automatic electrical bus transfer did not occur due to relay failures in the fast dead bus transfer system. Both 'A' and 'B' Emergency Diesel Generators started and loaded as designed to re-energize the 'A' and 'B' safety busses. The loss of off-site power caused a loss of both Main Feedwater pumps, resulting in an automatic actuation of the Emergency Feedwater system. The Root Cause of this event was the design change procedure used for modifications to the fast dead bus transfer circuitry did not include guidance to detect the susceptibility of the relays to DC coil inductive kick. The faulty relays in the fast bus transfer circuit were replaced prior to plant startup. An Unusual Event was declared at 1617 CDT due to loss of off-site power to safety buses for >15 minutes. All required safety-related equipment responded as expected during this event.
05000346/LER-2017-001	Davis Besse	20 July 2017 18 September 2017	Emergency Diesel Generator Fuel Oil Storage Tank Vents Not Adequately Protected from Tornado-Generated Missiles LER 17-001-00 for Davis-Besse, Unit 1, Regarding Emergency Diesel Generator Fuel Oil Storage Tank Vents Not Adequately Protected from Tornado-Generated Missiles	On July 20, 2017, with the Davis-Besse Nuclear Power Station (DBNPS) operating at approximately 100 percent power, it was identified that the Emergency Diesel Generator (EDG) fuel oil storage tank vents were not adequately protected from potential tornado-generated missiles. If a missile crimped the vent it could disable the transfer pump or tank, potentially impacting the seven-day fuel supply for the affected train(s) of EDG. While the storage tanks were protected from tornado missiles when installed, the vents were not provided with any such protection. Compensatory measures were established to ensure a vent path remained following a tornado event, and actions will be taken to ensure the vents for each EDG fuel oil storage tank are adequately protected from tornado missiles. This issue is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety, in accordance with 10 CFR 50.73(a)(2)(v) as a condition that could have prevented the fulfillment of the safety function, in accordance with 10 CFR 50.73(a)(2)(vii) as an event where a single cause or condition caused two independent trains to become inoperable in a single system, and in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
05000325/LER-2017-004	Brunswick Steam Electric Plant (Bsep) Unit1	17 September 2017	Emergency Diesel Generator and Primary Containment Isolation System Actuations	On September 17, 2017, at 0938 Eastern Daylight Time (EDT), a momentary power interruption to Emergency Bus E4 occurred during planned surveillance activities involving Emergency Diesel Generator (EDG) 4. This occurred when EDG 4 was disconnected from Emergency Bus E4 and offsite power was not supplying the bus. EDG 4 automatically transferred from manual mode to automatic control and reconnected to Emergency Bus E4. Normal frequency and voltage were restored with EDG 4 in automatic control. The momentary power interruption to Emergency Bus E4 resulted in various Unit 2 Primary Containment Isolation System (PCIS) actuations. The affected equipment responded as designed. The direct cause of this event was that Operators were not aware that, at the time of the event, Emergency Bus E4 was being supplied solely by EDG 4. As a result of a failed under-frequency relay, the incoming line and feeder breakers from Balance of Plant (BOP) Bus 2C to Emergency Bus E4 had opened during the performance of the EDG 4 surveillance, leaving only EDG 4 to power Emergency Bus E4 in the manual mode of operation.
05000334/LER-2017-002	Beaver Valley	19 July 2017 13 September 2017	1 OF 4 LER 17-002-00 for Beaver Valley Power Station, Unit No. 1 Regarding Inadequate Emergency Diesel Generator (EOG) Tornado Missile Protection Identified Due to Non-conforming Design Conditions	In order to address the concerns outlined in NRC Regulatory Issue Summary (RIS) 2015-06 "TORNADO MISSILE PROTECTION", evaluations of tornado missile vulnerabilities and their potential impact on Technical Specification (TS) plant equipment were conducted. This particular evaluation concluded that the following Structures, Systems, and Components (SSCs) are potentially vulnerable to tornado generated missiles: The Beaver Valley Power Station Unit 1 (BV-1) Emergency Diesel Generators (EDGs) engine exhaust piping is potentially vulnerable as a result of tornado generated missiles striking and subsequently crimping or crushing this piping rendering the EDGs inoperable. On July 19, 2017, both of the BV-1 TS required EDGs were declared inoperable and Enforcement Guidance Memorandum (EGM) 15-002 Rev 1 "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance," was applied. Compensatory measures were implemented within the time allowed by the applicable Limiting Condition(s) for Operation and both EDGs were then declared operable but nonconforming. The apparent cause of this issue was a lack of clarity during the original design and licensing of the plant that led to inadequate understanding of the tornado missile protection regulatory requirements. Actions will be taken to establish compliance for BV-1 EDGs either by a plant modification or employing a methodology for addressing tornado missile non- conformances for the EDG exhaust piping. This issue is reportable under 10 CFR 50.72 for a loss of safety function. However, enforcement discretion is being applied. As stated in EGM 15-002, Rev. 1, the NRC will exercise enforcement discretion for subsequent tornado missile 10 CFR 50.72 notifications. On February 23, 2017, FENOC provided the NRC the initial 10 CFR 50.72 notification in Event Notification (EN) number 52571 concerning tornado missile protection issues known at that time.
05000321/LER-2017-004	Hatch	24 August 2017	Tornado Missile Vulnerabilities Result in Condition Prohibited by Technical Specifications LER 17-004-00 for Edwin I. Hatch Regarding Safety Relief Valves' as Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria	On March 30, 2017 at approximately 0922 EST, with Unit 1 at 97 percent rated thermal power and Unit 2 at 100 percent rated thermal power, it was identified that based on the revised Enforcement Guidance Memorandum (EGM) 15-002, Revision 1, the Emergency Diesel Generator (EDG) fuel oil storage ventilation pipe extending approximately 5 feet above grade was reportable due to its nonconformance with tornado generated missile protection requirements. This condition caused the EDGs and their associated fuel oil storage system to be considered inoperable. However, due to implementing compensatory measures as required by EGM 15-002, the affected equipment was declared operable but non-conforming. These tornado missile vulnerabilities have existed since original plant construction and is a design legacy issue. Upon identification of the noncompliance, compensatory measures were taken to revise the abnormal procedure for natural occurring phenomena to include actions that must be taken for the diesel generator fuel oil storage tank vent lines following a tornado or high winds event. A design change is also being processed to prevent the vent lines from being impacted due to a tornado missile.
05000247/LER-2017-001	Indian Point	6 February 2017 22 August 2017	Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused By Main Boiler Feedwater Pump Turbine Low Pressure Governor Valves Failed Closed LER 17-001-00 for Indian Point, Unit 2 Regarding Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused By Main Boiler Feedwater Pump Turbine Low Pressure Governor Valves Failed Closed	On June 26, 2017, Operations commenced a downpower from 100 percent to 93 percent reactor power to support performance of the Main Turbine Stop and Control Valve Test. With reactor power at 94 percent, the 22 Main Boiler Feed Pump Turbine (MBFPT) speed control trouble alarm annunciated coincident with pump speed swings of 800 revolutions per minute (rpm). The operators ceased the downpower and placed the 22 Main Boiler Feedwater Pump (MBFP) in Manual speed control to control the rpm swings. This was unsuccessful, and the rpm swings continued. The 22 MBFPT low pressure (LP) governor valves were observed to be cycling from full-closed to full-open. The decision was made to take local pneumatic control of the 22 MBFP to stabilize pump speed. Two minutes after establishing local pneumatic control, the LP governor valves went to full closed. With the rapid reduction in 22 MBFP speed, the pump was no longer delivering feedwater flow to the SGs. An automatic main turbine runback signal should have been generated on a low speed signal; however, there was no turbine runback actuation. In response, the operators commenced a manual runback to reduce main turbine load, but the decreasing SG levels reached 15 percent, and at 1531 hours a manual reactor trip was initiated. All control rods fully inserted and all required safety systems functioned properly. The plant was stabilized in hot standby with decay heat being removed by the main condenser. The direct cause of the reactor trip was that the shoulder screws used on the 22 MBFPT LP governor valve servomotor linkage had backed out and detached. This caused the LP governor valves to fail closed, shutting off the turbine steam supply. This event had no effect on the public health and safety. The event was reported to the Nuclear Regulatory Commission (NRC) on June 26, 2017 under 10 CFR 50.72(b)(2)(iv)(B), 50.72(b)(2)(xi), and 50.72(b)(3)(iv)(A).
05000364/LER-2017-001	Farley	23 June 2017 21 August 2017	2B Emergency Diesel Generator Rendered Inoperable Due to a Jacket Water Leak without Makeup Capability LER 17-001-00 for Farley, Unit 2, Regarding 28 Emergency Diesel Generator Rendered Inoperable Due to a Jacket Water Leak without Makeup Capability	On June 23, 2017, during continued troubleshooting of a jacket water (JW) leak on the 2B Emergency Diesel Generator (EDG), it was determined that the backup service water (SW) makeup flow path to the JW expansion tank would not pass flow. This troubleshooting was conducted to validate operability assumptions made on April 21, 2017, after recurrence of a leak from the JW keep warm pump. Based upon the measured leak rate from the April 21st event, the 2B EDG would have been unable to meet its 7 day mission time without the use of makeup water to the JW expansion tank. Had the 2B EDG received a demand signal after March 3rd, the EDG may not have been able to perform its safety function during a design basis accident due to the JW leak rate and inability to makeup to the JW expansion tank. Since the 2B EDG may not have met its mission time from March 3, 2017 to April 21, 2017, the station unknowingly operated in a condition prohibited by Technical Specifications which is reportable under 50.73(a)(2)(i)(B).
05000335/LER-2016-003	Saint Lucie	21 August 2016 15 August 2017	Generator Lockout Relay Actuation During Power Ascension Results in Reactor Trip LER 16-003-01 for St. Lucie, Unit 1, Regarding Generator Lockout Relay Actuation During Power Ascension Results in Reactor Trip	On August 21, 2016, during Unit 1 restart following a maintenance outage, an unexpected actuation of the Main Generator Inadvertent Energization Lockout Relay caused the main generator to trip, resulting in an automatic reactor trip. The generator lockout prevented the automatic transfer of station auxiliaries to the available startup transformer power, requiring the emergency diesel generators to start and power the safety related buses. Reactor coolant pumps normally powered through the non-safety buses were deenergized, and decay heat removal was via natural circulation and Auxiliary Feedwater. The lockout relay actuation was caused by a latent error introduced during a 2013 design modification where a wire was inadvertently removed from the circuit. Corrective actions included restoration of the affected circuit and implementation of procedure guidance to verify the inadvertent energization relay state and to reset as required following Main Generator manual synchronization. This licensee event report is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) for system actuations of the reactor protection system, the emergency diesel generators and the auxiliary feedwater system. This event had no effect on the health and safety of the public.
05000461/LER-2017-008	Clinton	15 June 2017 11 August 2017	Division 3 Shutdown Service Water Pump Start Failure LER 17-008-00 for Clinton, Unit 1 re Division 3 Shutdown Service Water Pump Start Failure	On June 15, 2017, Clinton Power Station (CPS) commenced procedure CPS 9069.01, Shutdown Service Water Operability Test. The purpose of this procedure is to verify operability of the Division 3 Shutdown Service Water (SX) System Pump 1SX01PC and selected valves per the Inservice Testing program on a quarterly basis. At 0958, SX pump 1SX01PC was started and after approximately 30 seconds, it tripped due to thermal overload. The pump was declared inoperable and operations entered Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.7.2, Condition A which requires the High Pressure Core Spray (HPCS) system to be declared inoperable and enter TS LCO 3.5.1 Condition B which requires verification by administrative means that the Reactor Core Isolation Cooling (RCIC) system is operable and within 14 days restore the HPCS system to operable status. The cause of the event is under investigation. A supplemental report will be provided when the cause has been established. An ENS notification was made at 1214 (EN 52806). Because the HPCS system is a single train safety system, this event is reportable under 10 CFR 50.73(a)(2)(v)(D) as a condition that could have prevented the fulfilment of a safety function to mitigate the consequences of an accident.
05000313/LER-2017-002	Arkansas Nuclear Arkansas Nuclear One – Unit 1	27 May 2017 26 July 2017	High Pressure Injection Pump Inoperable for Greater Than Technical Specification Completion Time LER 17-002-00 for Arkansas Nuclear One, Unit 1, Regarding High Pressure Injection Pump Inoperable for Greater Than Technical Specification Completion Time	On May 27, 2017, an attempt to start the red train High Pressure Injection (HPI) pump (BJ) in accordance with normal operating procedures was initiated. Control Room operators received an annunciator, HPI PUMP TRIP, and observed no indication of the pump starting. During investigative walk downs with the relay department, personnel discovered the HPI breaker was not fully racked up (trip pedal still in a tripped (down) condition and roller not free to roll). Operations personnel performed manual breaker operations to rack the 4160 V breaker (EB) further in the up direction. The pump was successfully started and declared operable. The condition was the result of an inadequate risk evaluation. The associated Operations Directive has been revised to require Operations management approval when waiving start-checks of vital 4160 VAC components following racking up of the respective breaker. The revision is expected to ensure appropriate personnel are involved when determining the risk associated with not testing components for functionality/operability following racking up evolutions of an associated breaker.
05000341/LER-2017-003	Fermi	22 May 2017 21 July 2017	Division 2 Residual Heat Removal Service Water System (RHRSW) Inoperable Due to an Inoperable RHRSW Flow Control Valve LER 17-003-00 for Fermi 2 Regarding Division 2 Residual Heat Removal Service Water System (RHRSW) Inoperable Due to an Inoperable RHRSW Flow Control Valve	On May 22, 2017 at 05:10 am (EST), while placing Division 2 Residual Heat Removal Service Water (RHRSW) in service for biocide treatment of the Division 2 Residual Heat Removal (RI IR) Reservoir, the Division 2 RI IRSW Flow Control Valve (FCV) (El 1 50F068B) failed to fully open. Troubleshooting discovered the direct cause was failure of the anti-rotation bushing stem key. The apparent cause was system operating conditions (high vibration) resulting in the failed tack welds. Previous troubleshooting on an indication issue on May 5, 2017 for the RHRSW FCV was inadequate, and did not identify the failure of the anti-rotation key. As a result, the RHRSW FCV was returned to service at 2:50 pm on May 7, 2017, and subsequently failed on the next on-demand stroke at 5:10 am on May 22, 2017. Seventeen similar Motor Operated Valves (MOVs) were inspected and no MOVs exhibiting the symptoms observed on the E1150F068B prior to the failure of the anti-rotation key were found, and all anti-rotation devices were found to be intact. The Past Operability determination for 131150E068B found that the MOV was unable to perform its design basis functions from May 3. 2017 at 5:48 am, when the RI IRSW FCV was last successfully stroked under dynamic conditions, through May 24. 2017 at 4:04 pun, when the RI IRSW FCV was returned to service. The Division I RI-IRSW was available throughout the event except on two occasions. Division 1 of RHRSW was declared inoperable for Mechanical Draft Cooling Tower (MDCT) Nozzle Cleaning activities on May 9, 2017 from 8:41 am to May 9, 2017 at I I :18 pm. Division I of RI IRSW was again declared inoperable for IVIDCT Nozzle Cleaning activities on May 11, 2017 at 8:35 am through May 11, 2017 at 10:01 pm. The as found condition of the Division 2 RHRSW FCV is a condition prohibited by Technical Specification 3.7.1 and reportable under 10 CFR 50.73 (a)(2)(i)(13) "Operation or Condition Prohibited by Technical Specifications," and 10 CFR 50.73(a)(2)(v)(13) "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: Remove Residual Heat.
05000327/LER-2017-002	Sequoyah	14 July 2017	Automatic Actuation of Emergency Diesel Generators Due to Loss of Power to 6.9kV Shutdown Board LER 17-002-00 for Sequoyah, Unit 1, Regarding Automatic Actuation of Emergency Diesel Generators Due to Loss of Power to 6.9kV Shutdown Board	On May 23, 2017, at 2330 eastern daylight time (EDT), while transferring the 2A-A 6.9 kilovolt (kV) Shutdown Board (SDBD) from its alternate power source to its normal power source, in support of outage testing for Unit 2, a loss of power to the 2A-A 6.9kV SDBD resulted in the automatic actuation of all four emergency diesel generators. The power was restored to the SDBD on May 24, at 0027 EDT. At the time of the event, Unit 1 was in Mode 1 at 100 percent power and Unit 2 was in Mode 5 for a refueling outage. There were no actual safety consequences as a result of this event. The cause of the event was indeterminate. The direct cause of the event was an upstream unit board overcurrent relay actuation associated with the A phase. The corrective action was to obtain acceptable test results from all equipment in the zone of protection. This included all the cables, current transformers, relays and breakers that were in contact with the associated relay. There was no evidence of equipment degradation. Transient monitoring will be performed for 6.9kV SDBD normal and alternate feeder breakers. Data will be captured for each initial transfer. Any identified anomalies will be addressed via the corrective action program.
05000390/LER-2017-005	Watts Bar	10 May 2017 10 July 2017	Isolation of the 1 B-B Safety Injection Pump Leads to Condition Prohibited by Technical Specifications LER 17-005-00 for Watts Bar re Isolation of the 1B-B Safety Injection Pump Leads to a Condition Prohibited by Technical Specifications	On May 10, 2017, at 0907 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) Unit 1 operations personnel discovered the 1B-B Safety Injection pump discharge isolation valve (1-ISV-63-527) closed. Technical Specification (TS) 3.5.2, ECCS - Operating, Condition A was immediately entered for one or more trains of the Emergency Core Cooling System (ECCS) inoperable. TS 3.5.2 Condition A was exited at 0913 EDT when 1-ISV-63-527 was opened. Investigation determined that the 1 B-B SI pump discharge isolation valve had been closed prior to Unit 1 entering Mode 3 on April 26, 2017, representing a condition prohibited by TS. During this time period, the 1A-A SI pump was inoperable for 21 minutes, representing a condition that could have prevented fulfillment of a safety function. The cause of the mispositioned valve was the result of an individual failing to follow procedure use and adherence requirements during the performance of Emergency Diesel Generator (EDG) Blackout testing. The safety injection pump discharge valve was closed to support the test but was not reopened following the testing. Corrective actions for this event include personal accountability actions, revision of the EDG blackout procedures to ensure the SI pump discharge valves are reopened, and additional station focus on procedure use, particularly use of Not Applicable (N/A) in performing procedures.
05000461/LER-2017-002	Clinton	5 July 2017	Failure of the Division 1 Diesel Generator Ventilation Fan Load Sequence Relay Circuit During Concurrent Maintenance of RHR Division 2 Results in an Unanalyzed Condition LER 17-002-01 for Clinton, Unit 1, Regarding Failure of the Division 1 Diesel Generator Ventilation Fan Load Sequence Relay Circuit During Concurrent Maintenance of RHR Division 2 Results in an Unanalyzed Condition	On Tuesday, March 7, 2017 at 2258 CDT, an Equipment Operator detected a relay cycling every 10 seconds from the Division (Div.) 1 480V Unit Substation (Sub) 1A. Agastat Time Delay Relay (TDR) 427X2-41A (X2 relay) was determined to be the source of the clicking sound. The X2 relay supports the undervoltage load shed and restoration of the Div. 1 Emergency Diesel Generator (EDG) ventilation room fan 1VDO1CA. The logic is designed to shed 1VDO1CA on undervoltage, prevent it from restarting, then allow it to be restored 10 seconds after voltage is restored either by the EDG or return of the safety bus. An investigation of the relay cycling condition determined that 1VDO1CA was unable to respond to a demand signal and the cycling was the effect of interaction between the X2 and 427X3-41A (X3) relays. As a result, Div. 1 EDG was declared inoperable on March 9. With the Div. 2 Residual Heat Removal (RHR) System already inoperable due to scheduled maintenance, the plant was determined to be in an unanalyzed condition. A causal analysis determined that not understanding the design basis of the circuit subject to relay coordination and the impact of the change in specific components as part of a 2008 design change was the cause of the condition. Corrective actions included the replacement of the X3 relay with the original design relay. Div. 2 RHR was returned to service on March 8. Subsequently, Div. 1 EDG was restored to OPERABLE on March 11. This is reportable as an unanalyzed condition, operation prohibited by the Technical Specifications and a loss of safety function.
05000424/LER-2017-002	Vogtle	17 March 2017 29 June 2017	Vice President Vogtle Units 1-2 7821 River RD Waynesboro GA, 30830 706-848-0004 tel 706-848-3321 fax June 29, 2017 Docket No: 50-424 NL-17-1104 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Vogtle Electric Generating Plant — Unit 1 Licensee Event Report 2017-002-01 Wiring Error results in Automatic Actuation of a Safety System Ladies and Gentlemen: In accordance with the requirements of 10 CFR 50.73(a)(2)(iv)(A), Southern Nuclear Operating Company is submitting the enclosed Licensee Event Report, 2017-002-01 for Vogtle Electric Generating Plant Unit 1. This letter contains no NRC commitments. If you have any questions, please contact Dom Sutton at (706) 848-1428. Respectfully sub Darin J. M rs Vice President — Plant ogtle Units 1&2 - SNC DJM/KCW Enclosure: Unit 1 Licensee Event Report 2017-002-01 Cc: Regional Administrator NRR Project Manager — Vogtle 1 & 2 Senior Resident Inspector — Vogtle 1 & 2 RType: CVC7000 Vogtle Electric Generating Plant — Unit 1 Licensee Event Report 2017-002-01 Wiring Labeling Error results in Automatic Actuation of a Safety System Enclosure Unit 1 Licensee Event Report 2017-002-01 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (04-2017) LICENSEE EVENT REPORT (LER) ILWa t l,viricir (See Page 2 for required number of digits/charactersfor each block) ••••••• (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.qovireadincl-rmidoc-collections/nuregs/staff/sr1022/r3i) APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0313112020 Estimated burden per response to comply with this mandatory collection request: 80 hours Reported lessons learned are incorporated into the licensing process and fed back to industry Send comments regarding burden estimate to the Information Services Branch (T-2 F43), US Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects Resource@nrc gov, and to the Desk Officer, Office of Information •and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington. DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. 1. FACILITY NAME Vogtle Electric Generating Plant . DOCKET NUMBER 13. PAGE 17.:15000424 1 1 OF 2 4. TITLE Mirina Error results in_AulomaticActuation_of a Safety System_ LER 17-002-01 for Vogtle, Unit 1, Wiring Error Results in Automatic Actuation of a Safety System	On March 17, 2017, at approximately 1517 EDT, during refueling operations on Unit 1, power was being restored to the Open Phase system for the 1-B Reserve Auxiliary Transformer (RAT). During restoration, a valid undervoltage actuation signal was sent to the 1-B Emergency Diesel Generator (EDG). The EDG automatically started and tied to the safety bus. This undervoltage condition was caused by a wiring error in the Open Phase system to the 1-B RAT. Unit 1 was in Mode 6 at the time and remained so throughout the event. There was no change in the decay heat removal for the plant. Because this event resulted in the automatic actuation of a system listed in 10 CFR 50.73(a)(2)(iv)(B), this event is reportable under 10 CFR 50.73(a)(2)(iv)(A). This event had no adverse effect on the health and safety of the public, and is of very low safety significance. Unit 2 was unaffected.
05000425/LER-2017-001	Vogtle	29 June 2017	1 OF 3 LER 17-001-01 for Vogtle, Unit 2, Regarding Power Supply Failure Results in Operation in a Condition Prohibited by Technical Specifications Supplement	Between January 16 and February 13, 2017, a power supply that supports operation of one of two ventilation supply fans for the 2A Emergency Diesel Generator (EDG) failed. Vogtle Electric Generating Plant (VEGP) Technical Specifications (TS), 3.8.3 Condition F requires restoration of the fan within 14 days or the EDG must be declared inoperable. This condition was not identified by operators because there is no failure indication. Therefore, the power supply was not repaired within the required time. The EDG should have been declared inoperable and TS 3.8.1. Condition B entered. Subsequently the completion time for TS 3.8.1 Condition B expired and the unit should have been shut down. This action was not taken and Unit 2 operated in a condition prohibited by TS. Between February 13 and February 22, 2017, the second power supply failed on this EDG resulting in the second ventilation supply fan being inoperable. Since no failure indication is available, this condition was not identified by the operators. The TS 3.8.3 action for two power supplies inoperable is to immediately declare the EDG inoperable. Subsequently, the completion time expired and the unit operated in a condition prohibited by TS. During a monthly surveillance run on March 8, both power supplies were discovered failed. Both power supplies were replaced on March 9, 2017 and the EDG was declared operable.
05000368/LER-2017-002	Arkansas Nuclear Arkansas Nuclear One – Unit 2	26 April 2017 26 June 2017	Automatic Start of an Emergency Diesel Generator Due to the Momentary Loss of Offsite Power due to Severe Weather LER 17-002-00 for Arkansas Nuclear One, Unit 2, Regarding Automatic Start of an Emergency Diesel Generator Due to the Momentary Loss of Offsite Power due to Severe Weather	On April 26, 2017, ANO-2 was in day 28 of a refueling outage with the core completely off loaded to the spent fuel pool (SFP). Power to ANO-2 plant equipment was supplied from Start Up Transformer 2 (SU2) while SU3 was out of service for planned maintenance. 500kV and 161kV offsite power lines were in service. The area around the plant was experiencing severe weather from thunderstorms and tornado warnings had been issued from the National Weather Service for the four county area. Switchyard work was ceased. At approximately 1002 CST switchyard breakers for 500kV lines opened on fault current. High winds had damaged the transmission towers approximately 16 miles away from ANO and caused phase to ground faults. This resulted in a loss of all offsite power lines to the 500 kV bus. The autotransformer also locked out, as designed, when the 500 kV transmission lines faulted. When the 500kV bus tripped, the 4.16kV bus that feeds a vital 480 volt bus was subjected to a voltage transient; subsequently; the #1 emergency diesel generator (EDG) auto started. The EDG output breaker never closed due to the fact that voltage was restored to normal almost immediately. This EDG was secured due to running unloaded. Both SFP cooling pumps were out of service after the transient. A SFP cooling pump was restarted at 1020 CST. The temperature of the SFP did not change during this event.
05000313/LER-2017-001	Arkansas Nuclear Arkansas Nuclear One – Unit 1	26 April 2017 26 June 2017	Automatic Start of an Emergency Diesel Generator Due to the Loss of Offsite Power due to Severe Weather LER 17-001-00 for Arkansas Nuclear One, Unit 1, Regarding Automatic Start of an Emergency Diesel Generator Due to the Loss of Offsite Power due to Severe Weather	On April 26, 2017, Arkansas Nuclear One, Unit 1 (ANO-1), was operating normally at 100% rated thermal power. The 500kV transmission line to the substation at Pleasant Hill, Arkansas was out of service for planned maintenance. The area around the plant was experiencing severe weather from thunderstorms and tornado warnings had been issued from the National Weather Service for the four county area. At approximately 1002 CST switchyard breakers for 500kV lines opened on fault current. High winds had damaged the transmission towers approximately 16 miles away from ANO and caused phase to ground faults. This resulted in a loss of all offsite power lines to the 500kV bus. The autotransformer also locked out as designed when the 500kV transmission lines faulted. The Reactor Operator initiated a manual reactor trip about 8 seconds after the 500kV lines tripped and prior to the reactor protection system initiating an automatic trip. During this time both emergency diesel generators (EDGs) (EK) started as expected. EDG #2 re-energized one Engineered Safeguards bus. EDG #1 ran unloaded until shutdown. The plant was stabilized in Mode 3 with Emergency Feedwater (EFW) pumps supplying the steam generators, maintaining the water level at the natural circulation setpoint.
05000311/LER-2017-001	Salem	13 June 2017	Emergency Diesel Generator Start Due to a Loss of Power to the 2C 4160 Volt Vital Bus LER 17-001-00 for Salem, Unit 2, Regarding Emergency Diesel Generator Start Due to a Loss of Power to the 2C 4160 Volt Vital Bus	On April 14, 2017 at 13:57 while attempting to transfer the 2C 4 Kilovolt (kV) vital bus from 24 Station Power Transformer (SPT) to 23 SPT, the 24 SPT infeed breaker opened properly but 23 SPT infeed breaker failed to close. The 23 SPT infeed breaker failing to close as expected resulted in de-energization of the 2C 4kV vital bus and subsequent start and loading of the 2C Emergency Diesel Generator (EDG) to power the bus. This report is being made in accordance with 10CFR50.73 (a)(2)(iv)(A), "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)," for this event actuation of the 2C EDG. Notification of this event was provided via ENS report 52681 NS'C, FORM 366 I04-20 17:1
05000324/LER-2017-002	Brunswick	13 April 2017 12 June 2017	Plant Mode Change with Primary Containment Inoperable LER 17-002-00 for Brunswick, Unit 1, Regarding Foreign Material in Switch Results in Unplanned Automatic Start of Emergency Diesel Generators	On April 13, 2017, Unit 2 was in Mode 4 preparing to exit a refueling outage. The primary containment was being vented to ensure habitability of the Drywell. The valve alignment for Drywell ventilation makes the primary containment inoperable due to the Drywell and Suppression Chamber airspaces being in communication with each other. At 23:47 Eastern Daylight Time (EDT), the reactor mode was changed from Mode 4 to Mode 2 with ventilation still in progress. In Mode 2, the Primary Containment is required to be operable. Therefore, the plant entered a condition prohibited by the Technical Specifications, and the event is reportable per 10 CFR 50.73(a)(2)(i)(B). It is also reportable per 10 CFR 50.73(a)(2)(v)(D) because the primary containment safety function was lost. The condition was discovered 28 minutes later on April 14, 2017, at 00:15 EDT and was corrected by closing the ventilation flowpaths at 00:30 EDT on April 14, 2017. This event resulted from Control Room personnel not initiating a tracking document while in Mode 4 with the primary containment inoperable. When preparing to change the plant mode from Mode 4 to Mode 2, the primary containment ventilation status was overlooked. Corrective actions for this event included closing the containment ventilation paths and remediating the Shift Manager and Control Room Supervisor.
05000368/LER-2017-001	Arkansas Nuclear Arkansas Nuclear One – Unit 2	6 April 2017 30 May 2017	Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions LER 17-001-00 for Arkansas Nuclear One, Unit 2, Regarding Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions	On April 6, 2017, as part of the closure of an Arkansas Nuclear One, Units 1 and 2 (ANO-1 & 2) Tornado Protection Study, a nonconforming condition in the plant design for a conduit that contains safety related cables for the ANO-2 #1 Emergency Diesel Generator (EDG) meter and relay cabinets, was identified. The conduit did not meet current design basis for protection against a potential tornado missile impact. This vulnerability is similar to those previously reported in LERs associated with ANO-1. On April 6, 2017, Operations declared the #1 EDG inoperable, implemented Enforcement Guidance Memorandum (EGM) 15-002, “Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,” along with necessary compensatory measures, and subsequently declared the affected equipment operable but non- conforming. Interim corrections include implementation of compensatory strategies. Plant modifications and license basis changes are being evaluated to resolve outstanding issues. The cause of this issue was unclear and changing regulatory requirements during original plant licensing that led to an inadequate understanding of the regulatory guidance with respect to tornado missile protection design requirements.
05000255/LER-2017-001	Palisades	29 March 2017 24 May 2017	Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions LER 17-001-00 for Palisades Nuclear Plant Regarding Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions	On March 29, 2017, during an evaluation of protection of Technical Specification (TS) equipment from the damaging effects of tornados, nonconforming conditions were identified in the plant design. Specifically, TS equipment did not meet current design basis for protection against potential tornado missile impact. Identified components/systems were declared inoperable and NRC Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado Generated Missile Protection Noncompliance," was implemented. Initial compensatory measures were implemented, per the guidance of NRC Interim Staff Guidance DSS-ISG-2016-01 Appendix A, within the time allowed by the applicable Limiting Conditions for Operation (LCOs) and the associated systems were then declared operable but nonconforming. The six systems, containing TS required equipment, did not meet current design basis for protection against potential tornado missile impact. Credible tornado missile impacts could affect the following systems; Service Water, Fuel Oil, Emergency Diesel Generators, Auxiliary Feedwater, Component Cooling Water and Control Room Ventilation Filtration. Comprehensive compensatory measures will be implemented in approximately 60 days of discovery, per the guidance of NRC Interim Staff Guidance DSS-ISG-2016-01 Appendix A. Due to the historical nature of the issue, a specific cause for the identified vulnerabilities was not determined.
05000250/LER-2017-001	Turkey Point	18 March 2017 16 May 2017	Loss of 3A 4kV Vital Bus Results in Reactor Trip, Safety System Actuations and Loss of Safety Injection Function LER 17-001-00 for Turkey Point, Unit 3, Regarding Loss of 3A 4kV Vital Bus Results in Reactor Trip, Safety System Actuations, and Loss of Safety Injection Function	On March 18, 2017 at approximately 1107 hours, the Turkey Point Unit 3 reactor tripped from 100% power as a result of an electrical fault on the 3A 4kV vital bus. The Auxiliary Feed Water System actuated as expected, and the 3A Emergency Diesel Generator started but did not load, as designed, due to the lockout of the 3A 4kV bus. The 3A 4kV bus remained de-energized and the reactor was stabilized in Mode 3. Both Unit 4 High Head Safety Injection (HHSI) pumps were out of service for maintenance. The 3A HHSI pump was unable to be powered from the 3A 4kV bus resulting in a loss of the Safety Injection safety function for approximately 2.5 hours on both Units 3 and 4. The safety function is achieved by operation of two of the four pumps which are shared by both units. The loss of the 3A 4kV bus was caused by an electrical fault created by a conductive foreign material that had entered the current-limiting reactor cubicle that bridged an air gap between an uninsulated bus bar and the cubicle wall. The foreign material was a carbon fiber mesh used to reinforce a Thermo-Lag installation taking place in the 3A 4kV switchgear room. Corrective actions include: 1) The Thermo-Lag installation procedure will be revised to incorporate additional precautions for handling Thermo-Lag materials, and 2) the Engineering product risk and consequence assessment process will be revised to ensure a review is conducted of Safety Data Sheets for material being considered in the design. This event had no effect on the health and safety of the public.
05000389/LER-2017-002	Saint Lucie	15 May 2017	2A3 4.16 KV Bus De-Energization Due to Voltage Meter Failure	On May 15, 2017, at 1800 hours, the St. Lucie Unit 2 2A3 4.16 KV Bus undervoltage protection relays actuated resulting in a loss of power to the bus. The 2A emergency diesel generator (EDG) did not respond to this event as this EDG had been properly removed from service for pre-planned maintenance. The 2A3 4.16 KV Bus was restored to service at 2340 hours. The 2A3 4.16 KV Bus was de-energized when an internal fault within the 2A EDG local voltmeter blew fuses that removed power from the undervoltage relays that resulted in the loads powered from this bus being stripped. Corrective actions included replacing the fuses and replacing the susceptible local EDG voltmeter, as well as the interim use of caution tags on the susceptible voltmeter selection switches until modifications to remove the vulnerability are complete. During this event the B train safety related electrical busses remained operable and energized. All other equipment responded to the event per the existing plant conditions and the unit remained at 100% power. The A train safety related electrical bus was restored to service well within the Technical Specification allowed outage time. Therefore, this event had no significant impact on the health and safety of the public. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. Description On May 15, 2017, St. Lucie Unit 2 was in Mode 1 at 100 percent reactor power. The 2A emergency diesel generator (EDG) (EIIS:DG) was removed from service due to planned maintenance. At 1800 hours, the 2A3 4.16 KV Bus undervoltage protection relays (EIIS:27) actuated resulting in a loss of power to the bus (EIIS:SWGR). However, the 2A EDG did not respond to this event as this EDG had been properly removed from service for pre-planned maintenance. The troubleshooting team identified blown potential transformer (PT) fuses (EIIS:FU) in the 2A EDG metering circuit. The failed fuses were replaced and the 2A3 4.16 KV Bus was repowered at 2340 hours. The required NRC ENS notification for the system actuation was completed by 0017 hours on May 16, 2017. During this event the B train safety related electrical busses remained operable and energized and the unit remained at 100% power. Contents 1 Cause of the Event 2 Analysis of the Event 3 Safety Significance 4 Corrective Actions 5 Additional Information Cause of the Event The direct cause of the bus de-energization was determined to be failed secondary side PT fuses which provide power to the under voltage/degraded grid sensing circuity. The root cause was an internal failure within the local 2A EDG GE AB40 voltmeter (EIIS:MTR) causing a phase to phase short across the variable resistor. This condition resulted in the failure of the secondary PT fuses, resulting in the actuation of the 2A3 4.16 KV Bus UV relays. A contributing cause to this event was a latent design deficiency from original construction; the meter circuit did not have isolation fuses from the PT fuses. This allows an internal fault of the meter to open the protective circuit fuses and subsequently de-energize the UV relays. Analysis of the Event Even though the 2A EDG did not respond to the loss of the 2A3 bus because it was properly removed from service, the 2A3 4.16 KV bus UV protection relays did respond to the event and their actuation is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A). Two conditions contributed to the plant response to the event: 1. The local voltmeter selector switch was in the 1-3 position. This position allowed the meter fault to be simultaneously communicated to both phases of the PT fuses. These fuses provide sensing power to the UV relays which provide electrical isolation to the safety related 2A3 4.16 KV bus. Upon sensing the loss of power to the onsite power system, the safety portion of the system is automatically isolated from the non- safety portion of the system by the operation of circuit breakers on the lines between non-safety and safety related buses. 2. The 2A EDG was out of service for the on-line preventive maintenance period. This precluded the 2A EDG from starting and assuming the loads of the 2A3 4.16 KV bus. Section 8.3 of the UFSAR, Table 8.3-6 4.16 KV Safety Related System – Failure Modes and Effects Analysis describes the consequences of loss of offsite and EDG power to the 2A3 or 2B3 bus. The analysis states that the loss of an EDG in this case will result in the loss of a one safety related bus, however, the redundant safety system remains to supply the redundant safety related loads. Additionally, DC control power in this event remained unaffected, and the A side instrument inverters remained powered throughout by the A side DC battery. FPL performed extent of condition/extent of cause reviews for AC voltage metering circuits containing the GE AB40 voltmeter, as well as other voltmeters, for both units' safety related 4160 and 480 volt buses. The review determined that, with one exception, all remote voltage indication associated with the reactor turbine generator board (RTGB) voltmeters are all fused providing isolation of the metering circuit from their respective protective functions. The exception involves the Unit 2 local EDG voltmeters. The U2 EDG local metering circuits were determined to be the only safety related 4160 and 480 volt metering circuit whose failure could initiate a Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. protective function (e.g., blown PT fuses resulting in the de-energization of its associated bus UV relays). These reviews also identified several metering circuits that were not fuse protected; however, in these cases any proposed meter failure would only annunciate with no corresponding automatic protective function. Safety Significance The 2A EDG received a start signal from the under voltage condition on the 2A3 bus, but did not start as the EDG had been properly removed from service for preplanned maintenance. Upon a loss of indicated power to the potential transformers, the 2A3 4.16 KV bus responded appropriately for the existing plant conditions (e.g., the under voltage circuit relays actuated, the incoming breaker to the 4.16 KV bus opened, and a start signal was provided to the associated 2A EDG which was properly removed from service). Although the safety related 2A3 loads were lost during this event, the redundant loads serviced by the 2B3 train 4.16 KV safety related electrical bus remained unaffected by the event and the unit remained at 100% power. Normal power was restored to the 2A3 4.16KV bus within 6 hours of the event, well within the allowable 8-hour Technical Specification action statement for restoring the 2A3 4.16 KV bus. Therefore, this event had no significant impact on the health and safety of the public. The Unit 2 UFSAR section 8.3 describes Failure Modes and Effects for the 4.16 KV safety related system. This analysis bounds the observation of the event described in this LER. Corrective Actions EDG voltmeter selector switches directing that the switches not be left in the 1-3 position. 2. The local voltmeter for the 2A EDG was replaced. The following corrective action is being managed under the Corrective Action Program: 3. The local voltmeter for the 2B EDG will be replaced. 4. FPL is developing a modification to the 2A and 2B EDG metering circuit to install coordinated fuses between the metering circuit and the PT fuses to isolate the metering circuit from the UV relays in the event of a voltmeter fault. Failed Components Identified General Electric AB40 voltmeters Additional Information None

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