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Start date | Report date | Site | Reporting criterion | System | Event description | |
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05000289/LER-2017-004 | 5 February 2018 | Three Mile Island Three Mile Island Unit 1 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Emergency Diesel Generator | 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. | |
05000293/LER-2017-013 | 26 November 2017 25 January 2018 | 25 January 2018 | Pilgrim | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Secondary containment Primary containment Standby Gas Treatment System | On November 26, 2017, with the Reactor in the Run Mode at 100 percent power, while reviewing a procedure to be performed during normal scheduled testing it was determined that the test as written would cause both trains of Standby Gas Treatment System (SGTS) to be made inoperable during the test. This also made secondary containment system (SCS) inoperable. This LER is submitted to acknowledge that Pilgrim Nuclear Power Station missed providing Event Notifications and LERs for past occurrences. With both trains of SGTS and SCS inoperable while in Run, this event is reportable in accordance with Title 10 Code of Federal Regulations 50.73(a)(2)(v)(C) and 50.73(a)(2)(v)(D) as conditions that could have prevented the fulfillment of the safety function of a structure or system needed to control the release of radioactive material and mitigate the consequences of an accident. This has been determined to be a reportable condition that has not been reported during the past three years involving SGTS and secondary containment inoperability. The reportable conditions have occurred several times within the past three years during scheduled testing of SGTS. This event had no impact on the health and/or safety of the public. |
05000263/LER-2017-006 | 14 November 2017 12 January 2018 | 12 January 2018 | Monticello | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function | Reactor Coolant System Reactor Protection System Main Steam Isolation Valve Main Steam Line Main Steam | On November 14, 2017, it was identified that the use of the Reactor Protection System (RPS) test fixture described in some operations procedures would result in the loss of two RPS reactor Scram functions. Technical Specification 3.3.1.1 requires that RPS Instrumentation for Table 3.3.1.1-1 Function 5, Main Steam Isolation Valve-Closure and Function 8, Turbine Stop Valve-Closure, remain operable. It was concluded that a closure of three of four Main Steam Lines or Turbine Stop Valves would not necessarily have resulted in a full Scram during testing depending on the combination of closed valves occurring during the bypass condition. Operations procedures were revised to incorporate the use of the test fixture in December, 2008 for the Turbine Stop Valve Closure Scram Test Procedure and February, 2009 for the Main Steam Isolation Valve Closure Scram Test Procedure. The operations procedures were inappropriately revised to allow use of the test fixture on all RPS functions to prevent a half Scram. The operations procedures were quarantined until revisions were issued in December, 2017 that removed use of the test fixture. |
05000306/LER-2017-003 | 11 January 2018 | Prairie Island | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Service water Main Steam Isolation Valve Main Steam Line Containment Spray | On November 12, 2017 at 2119, a Control Room board walkdown discovered that both of the Unit 2 Containment Spray Pump control switches had been left in pull-out, when operators transitioned Unit 2 from Mode 5 to Mode 4. With the control switches in pull-out, the pumps would not automatically start as required. Technical Specification (Tech Specs) 3.0.3 was entered as a result of not complying with Technical Specification 3.6.5, Containment Spray and Cooling systems, which required both trains of Containment Spray to be Operable while in Mode 4. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Condition Prohibited by Technical Specification and 10 CFR 50.73(a)(2)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. The root cause determined that Surveillance Procedure SP 2099, Unit 2 Main Steam Isolation Valve Logic Test, was not adequately designed to account for outage schedule variation. Contributing causes included that the Unit 2 Startup to Mode 4 procedure does not contain adequate process barriers such that plant configuration meets Technical Specification requirements for Mode 4 entry. Operations personnel failed to uphold standards for panel walkdown requirements. Corrective actions include revising SP 2099, Unit 2 Main Steam Isolation Valve Logic Test to include steps to reposition Containment Spray Switches to the "as found" configuration and revise Unit 2 start-up procedure to add additional HOLD to have the Shift Manager perform Control Board Walkdown to verify equipment required in Mode 4 is aligned and Operable. Develop and implement an operations improvement plan specifically targeted to improve Operator standards in the performance of Control Board Walkdowns. | |
05000249/LER-2017-001 | 27 December 2017 | Dresden | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Reactor Pressure Vessel Standby Liquid Control Control Rod | Standby Liquid Control system subsystems were declared inoperable when control room personnel were notified of a through wall leak on the common discharge piping. Technical Specification (TS) 3.1.7, "Standby Liquid Control System," Condition B was entered. The pipe repair schedule projected that the work could not be completed within the allowed Completion Time of TS 3.1.7 and DNPS requested a Notice of Enforcement Discretion (NOED) to allow Unit 3 to remain at power during the repair. The NRC granted the NOED on September 12, 2017, at 1746 hours. The system was restored to operable status by replacing the piping on September 12, 2017, at 2035 hours within the time allowed by the NOED. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by TS. The cause of the event was a manufacturing defect. Corrective actions include replacing the failed piping (completed) and performing extent of condition pipe inspections. | |
05000390/LER-2017-016 | 20 December 2017 | 20 February 2018 | Watts Bar Nuclear Plant. Unit 1 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Emergency Diesel Generator Auxiliary Feedwater | 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. |
05000390/LER-2017-014 | 30 October 2017 20 December 2017 | 20 December 2017 | Watts Bar | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Ventilation | On October 30. 2017. at 0942 Eastern Daylight Time (EDT) Watts Bar Nuclear Plant (WBN) operations personnel received a Main Control Room (MCR) alarm for low control room positive pressure. At 0943 EDT, a Control Room Envelope (CRE) door was found ajar and immediately closed. Technical Specification 3.7.10 Control Room Emergency Ventilation System (CREVS) was declared not met for both trains, and Limiting Conditions for Operation (LCO) Condition B was entered for Unit 1 (Mode 1) and Condition G was entered for Unit 2 (Mode 5). At 0945 EDT the alarm cleared, CREVS was declared operable and LCO 3.7.10, Conditions B and G were exited. The loss of the control room envelope is being reported as a loss of safety function needed to mitigate the consequences of an accident. The cause of this issue is a human performance error in that an individual leaving the control building complex failed to confirm closure of the MCR envelope boundary door. Corrective actions have been generated to develop and install an engineering feature to inform personnel closing the door that it is fully shut and latched. |
05000440/LER-2017-006 | 1 December 2017 | Perry | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray | On October 4, 2017 at 0155 hours, while in Mode 1 at 100 percent rated thermal power, inoperability of both A and B trains of Motor Control Center, Switchgear, and Miscellaneous Electrical Equipment Areas Heating, Ventilation, and Air Conditioning System and Battery Rooms Exhaust System (M23/24) occurred. Train A was shutdown and declared inoperable based on excessive drive belt noise and belt malfunction. Train B was inoperable due to ongoing maintenance on its associated chilled water system. The combination of inoperability resulted in a loss of safety function. Technical Specification (TS) 3.0.3 was entered per plant procedures, and at 0250 hours a plant shutdown was commenced. At 0620 hours, the A train of M23/24 was declared operable following belt replacement and TS 3.0.3 was exited. The plant was restored to 100 percent rated thermal power at 0804 hours. The cause was determined to be inadequate procedural guidance in that the "general tensioning" method described in plant maintenance procedure, V-belt and Sheave Maintenance, is insufficient for restoring components to a reliable condition. Corrective action includes revising the procedure for correct tensioning guidance. The safety significance of this event is considered to be very small. This event is being reported in accordance with 10CFR50.73(a)(2)(v)(B), 10CFR50.73(a)(2)(v)(C), and 10CFR50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfilment of a safety function. NRC FORM 386 (06-2016) | |
05000397/LER-2017-007 | 3 October 2017 30 November 2017 | 30 November 2017 | Columbia | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Secondary containment Emergency Diesel Generator | 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. |
05000306/LER-2017-001 | 2 May 2016 29 November 2017 | 29 November 2017 | Prairie Island | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(b) 10 CFR 50.73(a)(2)(ii) | Main Steam Line Containment Spray | From May 2, 2016 to May 6, 2016, when B Train 122 Control Room Chiller (CRC) was out-of-service (OOS) per Technical Specifications (Tech Specs) 3.7.11 Condition A, Unit 2 A Train 23 Containment Fan Coil Unit (FCU) was OOS. According to revision 41 of site procedure C18.1, "Engineered Safeguards Equipment Support Systems," Bus 16 load sequencer and Bus 121 were inoperable when 122 CRC was OOS. Bus 121 supports B Train Diesel Driven Cooling Water Pump and Unit 2 B Train containment cooling (22/24 FCUs). So both trains of containment FCUs were OOS at the same time for approximately 35.6 hours. This would have required entry into LCO 3.0.3 putting Unit 2 in MODE 3 within 7 hours, this did not occur. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Operation or Condition Prohibited by Tech Specs. The cause was that the Senior Reactor operators failed to utilize Human Performance Tools (Verification/Validation and Procedure Use/Adherence) when assessing the Tech Specs impact to Unit 2 for applying LCO 3.0.6 when 122 CRC was taken OOS. Corrective actions include independent assessment of shared system LCO's for each unit, revising the LCO database, established a standard for LCO 3.0.6 log entries, and revising the safety function determination program to be more user friendly. |
05000254/LER-2017-003 | 21 September 2017 17 November 2017 | 17 November 2017 | Quad Cities | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | HVAC Control Room Emergency Ventilation | On 09/21/2017 at 1550, Operations started Control Room Emergency Ventilation (CREV) Air Conditioning (AC) for Tracer Gas Testing. Per Operations instructions, Mechanical Maintenance went to inspect the Control Room Emergency Ventilation system for refrigerant leaks before Tracer Gas Testing was started. Mechanical Maintenance reported a refrigerant leak on the discharge piping of the compressor, right above the inlet to the condenser. The leak was at the expansion joint of a fitting. The safety significance of this event was minimal. The cause of the refrigerant leak on the Control Room Emergency Ventilation compressor discharge pipe fitting into the condenser was due to high cycle fatigue. The corrective action was to replace the failed Control Room Emergency Ventilation compressor discharge pipe fitting. The CREV AC system is a single train system. Given the impact on the CREV AC system, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of any event or 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. |
05000266/LER-2017-001 | 18 September 2017 16 November 2017 | 16 November 2017 | Point Beach | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | At 1724 (CDT) on 9/18/17 Door-061, South Control Room Door (DR) was inadvertently disabled. The door became wedged open against its backstop during control room ventilation testing. Door-061 is a barrier that functions to maintain the control room envelope (NA). The barrier was subsequently disengaged from the backstop allowing it to close. The door was inspected and returned to operable status at 1750 (CDT). While the door was stuck open, the control room was in an unanalyzed condition, a condition that could have prevented fulfillment of a safety function, and a common cause inoperability of independent trains or channels. This event is being reported pursuant to 10 CFR 50.73(a)(2)(ii)(B), 10 CFR 50.73(a)(2)(v)(A), 10 CFR 50.73(a)(2)(v)(D), and 10 CFR 50.73(a)(2)(vii) for a degraded barrier that affected the control room envelope. | |
05000397/LER-2017-005 | 12 September 2017 9 November 2017 | 9 November 2017 | Columbia | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Secondary containment Reactor Core Isolation Cooling | On September 12, 2017 at 1227 PDT, Secondary Containment became inoperable due to pressure increasing above the Technical Specification limit of -0.25 inches of water gauge. While the plant was at 100% power, a Reactor Building exhaust valve and supply valve unexpectedly lost power and closed, resulting in a loss of Secondary Containment for approximately one minute. While Technical Specification limits were exceeded for this short time period, the resulting pressure excursion was bounded by analytical results; and thus, there were no safety consequences for this condition. This event was reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D) as Event Notification #52966. The apparent cause of the event was that station personnel did not deliberately and conservatively perform work tasks. Workers failed to update work instructions when work was rescheduled, and did not verify power sources at the work site. Corrective actions for this event include conducting a workshop on management expectations of Maintenance, increased management oversight, and addressing human performance issues. |
05000341/LER-2017-005 | 3 November 2017 | Fermi | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Feedwater Service water High Pressure Coolant Injection Emergency Diesel Generator Reactor Core Isolation Cooling Residual Heat Removal Automatic Depressurization System Emergency Equipment Cooling Water | 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-2017-010 | 17 August 2017 10 October 2017 | 16 October 2017 | Watts Bar | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Auxiliary Feedwater Emergency Core Cooling System Control Rod | On August 17, 2017, at 1205 Eastern Daylight Time (EDT), the Watts Bar Nuclear Plant (WBN) lost power to the 1B-B 6.9kV Shutdown Board. The loss of power to this safety related bus resulted in an automatic start of the Unit 1 Turbine Driven Auxiliary Feedwater Pump (TDAFWP). Power to the 1B-B Shutdown Board (SDBD) was restored at 1505 EDT on August 17, 2017. During the loss of power to the 1B-B SDBD, a reduction in containment and control rod drive mechanism cooling occurred. At 1233 EDT, lower containment average temperature exceeded Technical Specification (TS) limits, and TS 3.6.5 Condition A was entered for containment average air temperature not within limits. Lower containment average temperature was restored to within limits at 1525 EDT on August 17. 2017. This is reportable as a potential loss of safety function. The cause of this event is mechanical vibration while closing a panel drawer resulting in actuation of protective relays that led to a loss of power. Clearances will require the relays involved in this event to be isolated during drawer movement to prevent a similar occurrence. |
05000440/LER-2017-004 | 8 August 2017 4 October 2017 | 4 October 2017 | Perry | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray Emergency Core Cooling System | On August 8, 2017, at 1554 hours, while the plant was at 100 percent rated thermal power, during restoration from testing of the High Pressure Core Spray (HPCS) Suppression Pool (SP) Level High Instrumentation, unexpected as-left indications were found that impacted both of the required channels of instrumentation. With both SP level instruments inoperable, a loss of safety function existed. While venting the sensing line, the HPCS system was aligned to the suppression pool water source. This source of water is HPCS's safety-related source of water. The automatic suction swap on high suppression pool level is implicitly assumed in the accident and transient analysis since it assumes that the HPCS suction source is the suppression pool. Since the HPCS system was aligned to the suppression pool when the failure occurred, the assumptions of the accident analysis are met, and no safety system functional failure occurred. The cause for the unexpected as-left indications is due to air entrained in the sensing line which came out of solution. The safety significance of this event is considered to be small. This event is being reported under 50.73(a)(2)(v)(D) for a loss of safety function. |
05000440/LER-2017-005 | 14 August 2017 4 October 2017 | 4 October 2017 | Perry | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Secondary containment | On August 14, 2017 at 2257, with both trains of the Annulus Exhaust Gas Treatment System (AEGTS) running in parallel, Secondary Containment vacuum momentarily degraded to 0.52" water gauge when AEGTS B was shutdown to Standby Readiness. After an approximately 15 second delay, AEGTS A responded to maintain the proper vacuum. Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.1 requires a minimum of 0.66" water gauge differential pressure to be maintained between ambient and annulus pressure. The failure to comply with SR 3.6.4.1.1 caused Secondary Containment to be inoperable resulting in Technical Specification 3.6.4.1 not being met. The cause was determined to be equipment age-related degradation of the AEGIS A controller circuit card, which was subsequently replaced. There were no other event-related equipment malfunctions. Since AEGTS is not a core damage mitigation system and is not credited in the PRA model to mitigate large and early containment releases, the inoperability of the AEGTS is determined to be of small safety significance. This event is being reported in accordance with 10CFR50.73(a)(2)(v)(C) and 10CFR50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfilment of a safety function. |
05000397/LER-2016-002 | 27 September 2017 | Columbia | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Secondary containment Standby Gas Treatment System | On October 3, 2016 at 1008 PDT, the Secondary Containment (Reactor Building) became inoperable due to p:'essure increasing above the Technical Specification limit of -0.25 inches of water gauge (inwg). While the plant was at 100% power, a Reactor Building exhaust valve (REA-V-1) unexpectedly closed, resulting in a loss of Secondary Containment vacuum for approximately four minutes. Operations personnel manually started the Standby Gas Treatment System A and quickly restored Secondary Containment to less than -0.25 inwg. While Technical Specification limits were exceeded for this short time period, the resulting pressure excursion was bounded by analytical results; and thus, there were no safety consequences for this condition. This event was reported under reporting criterion 10 CFR 50.72(b)(3)(v)(C) as Event Notification #52276. The cause of the REA-V-1 closure is currently under investigation; corrective actions for this condition will be determined upon completion of the in% estigation. NRC FORM 386 (06-2016) | |
05000387/LER-2017-006 | 6 September 2017 | Susquehanna | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Secondary containment | On October 6, 2017 at 1945 hours, a loss of Control Room Habitability Envelope (CRHE) was declared due to failing to meet Technical Specification (TS) 3.7.3, Surveillance Requirement (SR) 3.7.3.4 during surveillance testing. The CRHE is required to be maintained such that occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. Mitigating actions were identified and include the ability to issue potassium iodide (KI) to the control room staff and Emergency Plan responders. With implementation of the mitigating actions, the dose consequence to the control room operators and affected Emergency Plan personnel is substantially less than the regulatory limit of 5 Rem Total Effective Dose Equivalent (TEDE) for event duration. On October 6, 2017, at 2146 hours, this condition was reported (ENS #53003) in accordance with 10 CFR 50.72(b)(3)(v)(D) and is also reportable in accordance with the corresponding criteria of 10 CFR 50.73(a)(2)(v)(D). Additional review resulted in an assumption that the gasket could have been out of position since installation in 2012; therefore, this condition is also being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications. The cause was out of position door gaskets on the "A" Control Room Emergency Outside Air Supply (CREOAS) filter train due to inadequate installation. Corrective actions include repairing the door gaskets, performing a successful retest, and adding a step in the preventive maintenance activity when replacing the door gaskets to ensure the gasket is entirely glued into the channel such that the gasket is in contact with the entire channel. There were no actual consequences to the health and safety of the public as a result of this event. | |
05000397/LER-2016-003 | 20 November 2016 29 August 2017 | 12 January 2017 | Columbia | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Secondary containment HVAC | On November 20, 2016 at 1402 PST, Secondary Containment (NH) (Reactor Building) became inoperable due to pressure increasing above the Technical Specification limit of -0.25 inches of water gauge (inwg). While the plant was ascending in power, the Reactor Building exhaust air fan unexpectedly failed to start in manual during post-maintenance testing. Prior to this event, Reactor Building Heating, Ventilation and Air Conditioning (VA) (HVAC) System A was running. Per station procedures, System A was stopped and System B was to start. The fan's failure to start resulted in no Reactor Building fans running, and increased Reactor Building pressure. For a time period of less than one minute, Secondary Containment pressure was not maintained less than or equal to -0.25 inwg. Immediate recovery actions by Operations personnel included manually starting Reactor Building HVAC System A, which quickly restored Secondary Containment pressure to less than or equal to -0.25 inwg at 1403 PST. While TS limits were exceeded for this short time period, the resulting pressure excursion was bounded by analytical results; and thus, there were no safety consequences for this condition. This event was reported under reporting criteria 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D) as Event Notification #52382. The cause of the exhaust fan's failure to start was a faulty control switch for the fan. Corrective actions for this event include replacement of the control switch. There were no other event-related equipment malfunctions. |
05000321/LER-2017-004 | 24 August 2017 | Hatch | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Emergency Diesel Generator Safety Relief Valve | 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. | |
05000397/LER-2017-003 | 26 June 2017 24 August 2017 | 24 August 2017 | Columbia | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Secondary containment HVAC | On June 6, 2017 at 1756 PDT hours Secondary Containment pressure exceeded the Technical Specification (TS) limit during a period of inclement weather. At 1756 PDT Secondary Containment was declared inoperable and operations personnel entered TS Action Statement 3.6.4. I .A and subsequently exited at 1800 PDT. Secondary Containment pressure was restored automatically by system response and operator action was not required. The direct cause of the momentary loss of Secondary Containment was due to slow system response to maintain a vacuum in Secondary Containment during a period of inclement weather. The interim planned corrective action is to verify proper operation and tuning of the Secondary Containment instrumentation. Additionally Columbia Generating Station is pursuing the change to TS requirements by adopting TSTF-551, Revise Secondary Containment Surveillance Requirements. This condition is being reported under 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and to mitigate the consequences of an accident. |
05000373/LER-2017-007 | 22 June 2017 18 August 2017 | 18 August 2017 | Lasalle LaSalle | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Reactor Pressure Vessel Core Spray Reactor Building Ventilation Residual Heat Removal Emergency Core Cooling System | On June 22, 2017, the Unit 1 Low Pressure Core Spray (LPCS) system was declared inoperable due to loss of corner room area cooling and loss of motor cooling. The common diesel generator cooling water pump received an automatic trip signal while being secured. The LPCS pump remained in standby during the event. Troubleshooting identified the most likely cause for the trip of the emergency cooling water pump breaker was a malfunction of the cooling water pump control switch or the cooling water supply fan control relay. Both suspected components were replaced. The causal investigation did not identify a specific cause; however, there is a high level of confidence that the failure modes were eliminated by the corrective actions taken during troubleshooting. This component inoperability is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. This condition could have prevented the LPCS system, a single train safety system, from performing its design function. There was minimal safety consequences associated with the condition since other required emergency safety systems remained operable, there were no actual demands for Unit 1 LPCS, and safety margins were maintained. |
05000298/LER-2017-004 | 17 August 2017 | Cooper | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Primary containment | On June 19, 2017, during performance of surveillance 6.PC.207, "Torus to Drywell Vacuum Breaker Operation," the control switch for vacuum breaker PC-AOV-NRV21 was cycled open, then closed. When the control switch was taken to close, the vacuum breaker failed to indicate closed. As such, Operations declared primary containment and PC-AOV-NRV21 inoperable and entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.1 Condition A and LCO 3.6.1.8 Condition B at 21:15 hours. In addition, TS LCO 3.6.1.1 Condition B was entered at 22:15 hours due to PC-AOV-NRV21 still indicating intermediate. The control switch for PC-AOV-NRV21 was cycled open, and then closed a second time. At this time, PC-A0V- NRV21 indicated closed. Operations declared primary containment and PC-A0V-NRV21 operable at 23:11 hours and exited TS LCO 3.6.1.1, Condition A and Condition B, and TS LCO 3.6.1.8, Condition B. The cause is currently under investigation. Cooper Nuclear Station will provide a supplement to this Licensee Event Report after inspection of the vacuum breakers can be performed (i.e., during Refueling Outage 30). There were no safety consequences associated with this condition. | |
05000263/LER-2017-004 | 19 June 2017 16 August 2017 | 16 August 2017 | Monticello | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection Reactor Core Isolation Cooling Automatic Depressurization System | On June 19, 2017 following a planned High Pressure Coolant Injection (HPCI) system maintenance, a HPCI start attempt was performed per the quarterly test procedure. HPCI failed to start during the test due to the steam stop valve HO-7 not opening caused by HO-7 oil relay not functioning properly. Since the component was not the subject of the maintenance activity, the HPCI failure was reported to the NRC under Emergency Notification System, Event Number 52814. The HO-7 oil relay was repaired and the HPCI system was returned to operable status at 13:30 on June 23, 2017. |
05000461/LER-2017-008 | 15 June 2017 11 August 2017 | 11 August 2017 | Clinton | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Service water Emergency Diesel Generator Reactor Core Isolation Cooling High Pressure Core Spray | 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. |
05000286/LER-2017-002 | 11 June 2017 9 August 2017 | 9 August 2017 | Indian Point Docket Number | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function | Reactor Coolant System Feedwater Residual Heat Removal Decay Heat Removal | On June 11, 2017, while at 100 percent reactor power, Operations placed Chemical and Volume Control System (CVCS) Demineralizer Diversion Valve CH-TCV-149 in DIVERT to allow the 32 Mixed Bed Demineralizer to be removed from service and align the 31 Mixed Bed Demineralizer for service. Within about two minutes after returning CH-TCV-149 to AUTO, which placed the 31 Mixed Bed Demineralizer in service, Letdown Backpressure Control Valve CH-PCV-135 demand had gone to 0 percent (full open demand) while letdown backpressure had increased, reaching 302 psig. Operations was alerted to a leak that had developed on 32 Mixed Bed Demineralizer Inlet Isolation Valve CH-352. In an effort to isolate the leak, CH-TCV-149 was placed in DIVERT. Due to the elevated pressure at CH-TCV-149 with CH-PCV- 135 fully open, placing CH-TCV-149 in DIVERT coupled with the elevated line pressure created a pressure transient in the letdown line upstream of the CVCS Reactor Coolant Filter. Reactor Coolant Filter Inlet Isolation Valve CH-305 experienced this pressure transient, which resulted in the valve developing a significant leak at the body to bonnet joint. Abnormal Operating Procedure (AOP) 3-AOP-LEAK-1 was entered, and normal letdown was manually isolated to stop the CH-305 leak. Excess letdown was placed in service to balance reactor coolant inventory at a Pressurizer water level of 61 percent. This exceeded the 54.3 percent limit of Technical Specification 3.4.9 Condition A, and Operations declared the Pressurizer inoperable. The inoperability of the Pressurizer is reportable as a safety system functional failure under 10 CFR 50.73(a)(2)(v). The direct cause of this event was elevated system pressure due to loading of the Reactor Coolant Filter from materials when the 31 Mixed Bed Demineralizer pathway was aligned. The elevated operating pressure in the CVCS letdown stream challenged the integrity of diaphragm valves CH-352 and CH-305, requiring the isolation of normal letdown. |
05000374/LER-2017-003 | 11 February 2017 9 August 2017 | 9 August 2017 | Lasalle LaSalle | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Reactor Pressure Vessel High Pressure Core Spray Emergency Core Cooling System Main Steam Line Main Steam | On February 11, 2017, Unit 2 was in Mode 5 for a planned refueling outage. While attempting to fill and vent the Unit 2 High Pressure Core Spray (HPCS) system, no flow was observed from the drywell vent valves or downstream of the HPCS injection valve. The HPCS system was already inoperable to support scheduled surveillances performed on February 8, 2017 in which the HPCS injection isolation valve had been cycled five times satisfactorily. Troubleshooting determined the cause of the valve malfunction was due to stem-disc separation. The valve internal components were replaced prior to restart of the unit from the refueling outage. The root cause of the valve failure was insufficient capacity of the shrink-fit stem collar, combined with multiple high-load cycles, which resulted in loosening and eventual shear failure of the wedge pin and threads. This component failure is reported in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. This condition could have prevented the HPCS system, a single train safety system, from performing its design function if the valve failure occurred during an actual demand. This component failure is also reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications (TS) 3.5.1 "ECCS - Operating," since the HPCS system could have been 1 inoperable for greater than the TS 3.5.1, Required Action B.2, Completion Time of 14 days to restore HPCS system to operable status. There were minimal safety consequences associated with the condition since HPCS was not required to be operable at the time of the failure, and other required emergency safety systems remained operable. There were no actual demands for Unit 2 LHPCS, other ECCS systems, or the reactor core isolation cooling (RCIC) system during this period. - --- ------- - NRC FORM 366 (04-2017) - 01 003 2017 |
05000325/LER-2017-003 | 5 June 2017 2 August 2017 | 2 August 2017 | Brunswick | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Ventilation Control Room Envelope Control Rod | On June 5, 2017, at 0930 Eastern Daylight Time (EDT), Unit 1 was in Mode 1 at 100 percent power, and Unit 2 was in Mode 1 at 87 percent power and was increasing power after a preplanned control rod improvement evolution. Maintenance personnel were inspecting dampers in the Control Room Air Conditioning (AC) system and Control Room Emergency Ventilation (CREV) system and disconnected an air supply to a damper. This resulted in the CREV system being inoperable due to interruption of the pneumatic supply. The CREV system was restored to operable status by 1009 EDT. At 1352 EDT, a second damper was inspected, and its pneumatic supply was disconnected. During the second occurrence, the Control Room AC system also tripped and was made inoperable. Affected systems were restored by 1407 EDT. The event is reportable as a loss of safety function per 10 CFR 50.73(a)(2)(v)(D). The event resulted from inadequate use of human performance tools and inadequate work instructions. Corrective actions for this event include restoring the affected pneumatic supply, revising work orders, and taking steps to emphasize proper use of human performance tools. |
05000298/LER-2017-003 | 24 July 2017 | Cooper | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | On May 26, 2017, a Control Room Emergency Filter System (CREFS) supply fan damper was discovered to be partially open, limiting air flow to less than the Technical Specifications (TS) required flow. The configuration of the supply fan damper is to be fully open when the corresponding supply fan is in service. Cooper Nuclear Station (CNS) Operations personnel declared CREFS inoperable at 1105 and entered TS Limiting Condition for Operation 3.7.4, Condition A. Subsequent investigation revealed that the damper control arm was correctly positioned, however the T-handle for the control arm was overtightened causing the control arm to bend upward which mispositioned the damper from full open to partially open. The control arm for HV-AD-AD1021B was replaced and post work testing performed to ensure the correct position of the damper. CREFS was declared operable at 0432 on May 27, 2017. To prevent recurrence, CNS will modify the design of the both CREFS damper control arms and the means of securing in position to prevent control arm bending. There were no safety consequences associated with this condition. | ||
05000293/LER-2017-001 | 16 January 2017 17 July 2017 | 17 July 2017 | Pilgrim | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Secondary containment HVAC Reactor Pressure Vessel Standby Gas Treatment System Core Standby Cooling System | Station (PNPS) was performing surveillance testing of secondary containment isolation dampers when dampers AO-N-82 and AO-N-83, refueling floor supply isolation dampers, failed to fully close when the control switches were taken to close. The failure of dampers AO-N-82 and AO-N-83 to fully close resulted in a loss of safety function for secondary containment, causing immediate entry into Limiting Condition for Operation (LCO) Action Statement (AS) 3.7.C.2.a, at 1155 hours. This LCO AS was exited at 1206 hours when the dampers were verified closed. An 8-hour non-emergency notification was made in accordance with 10 CFR 50.72(b)(3)(v), any event or condition that at the time discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. In addition, this notification is being conservatively made by PNPS in accordance with 10 CFR 50.73(a)(2)(i)(B), as a condition that was prohibited by Technical Specifications. The reactor building isolation dampers were cleaned, lubricated and post-work tested. PNPS has returned the dampers to operable status. Planned action to prevent recurrence is to revise the preventive maintenance strategy. There was no impact to public health and safety from this condition. |
05000373/LER-2017-006 | 17 May 2017 14 July 2017 | 17 July 2017 | Lasalle LaSalle | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Reactor Core Isolation Cooling Core Spray Residual Heat Removal Emergency Core Cooling System Main Steam | On May 17, 2017 at 0908 CDT, during Unit 1 full-power operations, operators received an unexpected alarm for the Low Pressure Core Spray (LPCS) pump injection high flow and automatic closure of the LPCS minimum flow valve (1E21-F011). Inspections indicated the flow switch that actively controls the LPCS minimum flow valve had a faulty diaphragm which allowed for water intrusion into the device. There were no impacts on plant operations. The required actions of Technical Specifications 3.5.1, "ECCS - Operating" and TS 3.3.5.1, "Emergency Core Cooling System (ECCS) Instrumentation" were entered. The switch was replaced and LPCS system tested, which allowed full restoration of the system on May 17, 2017 at 18:45 CDT. This condition could have prevented the LPCS system from performing its design function. This condition is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. There was minimal safety consequences associated with the condition since other emergency safety systems remained operable. |
05000265/LER-2017-001 | 15 May 2017 13 July 2017 | 14 July 2017 | Quad Cities | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Feedwater High Pressure Coolant Injection Primary containment Core Spray Low Pressure Coolant Injection | On May 15, 2017, at 19:18 hours, station Operations personnel were performing a High Pressure Coolant Injection (HPCI) Pump Operability Test which ensures the HPCI Minimum Flow Valve opens as pump flow decreases. When the HPCI Turbine was tripped, the Minimum Flow Valve did not open as expected when system flow was reduced to the low flow setpoint. Operators took steps to open the valve manually, but upon release of the control switch, the valve returned to the closed position. The valve was then left in the closed position. The HPCI system was declared inoperable and Technical Specification 3.5.1 Condition G was entered. The cause of the Minimum Flow Valve failing to open was attributed to the HPCI Pump Discharge Flow Indicating Switch, specifically, intermittent failure of the high side micro switch caused by residual material from the manufacturing process. The Flow Indicating Switch, which had been installed for three months, was replaced and the HPCI Pump Operability Test was successfully re-performed. The failed switch was then sent to Exelon's Power Labs for failure analysis. The safety significance of this event was minimal. Given the impact on the HPCI system, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of any event or 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. The HPCI system is a single train system and the loss of HPCI could impact the plant's ability to mitigate the consequences of an accident. |
05000458/LER-2017-006 | 15 May 2017 13 July 2017 | 13 July 2017 | River Bend | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Core Spray | On May 15, 2017, an engineering investigation determined that a modification installed in 2014 on two of the four safety-related main control building chillers had a design error. The nature of that error was such that the performance of a regularly scheduled preventative maintenance (PM) task to draw an oil sample from'the chiller gearbox inadvertently caused the chiller to be incapable of responding to an automatic start signal. A review of the history of the PM found that, on three occasions since the modification was installed, the task was performed on the operable chiller that was in the standby condition. The inadvertent inoperability of the standby division of the main control building chillers causes the loss of safety function of the supported electrical distribution systems in the building. The control building chilled water system provides cooling to the equipment rooms housing the battery chargers and inverters for the safety-related onsite electrical distribution systems. The loss of cooling to the various equipment rooms in the control building requires that the supported equipment in those areas be declared inoperable. The Technical Specifications for the Division 3 DC distribution system requires that the high pressure core spray (HPCS) system be immediately declared inoperable. This condition potentially causes the HPCS system to be incapable of performing its safety function, and is, thus, reportable in accordance with 10 CFR 50.73(a)(2)(v)(D). The error in the subject modification is considered a legacy issue since its design was completed and approved in July 2012. The PM task will be revised to preclude its performance on chillers in the standby configuration. At no time during the three performances of the PM on the operable standby chiller was there an actual demand for its automatic start. This condition was, thus, of minimal significance with respect to the health and safety of the public. I |
05000286/LER-2017-001 | 14 May 2017 13 July 2017 | 13 July 2017 | Indian Point Indian Point Unit 3 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | Reactor Coolant System Residual Heat Removal Emergency Core Cooling System | On May 14, 2017, at 0233 hrs, Indian Point Unit 3 entered Mode 4 as part of coming out of outage 3R19 and preparing for power operations. Operations test group was preparing for performance of 3-PT- CS004, Residual Heat Removal (RHR) Check Valve Testing. The team gathered for a pre job brief in accordance with the requirements of EN-HU-102, Human Performance Traps &Tools Procedure. At the time the only allowable access point to the Inner Crane Wall was through the double gate combination of Gates D and E, which require one gate to be maintained closed and secured at all times. Workers needed to enter inside of the Crane Wall to perform a portion of the valve lineup required by 3-PT-CS004. After unbolting and opening the gate, the two operators and a contract Radiation Protection (RP) Technician went through gate C despite a posted sign stating that the gate was not to be utilized in modes 1 through 4. While the valve manipulations were in progress the NRC Resident Inspector was also conducting a tour of the Vapor Containment (VC) and identified that gate C was opened. This gate being open in this plant condition resulted in a safety system functional failure, since with the gate unsecured this made the containment sumps inoperable. |
05000390/LER-2017-005 | 10 May 2017 10 July 2017 | 10 July 2017 | Watts Bar | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Reactor Coolant System Emergency Diesel Generator Emergency Core Cooling System | 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. |
05000416/LER-2017-005 | 9 July 2017 | 7 September 2017 | Grand Gulf | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | On July 9, 2017, at approximately 2158 hours central daylight time a pressure boundary door to the Control Room Envelope was left unsecure. The Control Room Envelope was inoperable for approximately one minute, at which time the door was closed. Grand Gulf Nuclear Station personnel identified that a loss of Safety Function occurred due to a breach in the Control Room Envelope resulting in inoperability of both divisions of Standby Fresh Air. This event is being reported under 10CFR 50.72(a)(2)(v)(D), as any event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. The cause was determined to be that the organization failed to understand the nuclear safety consequence associated with a degraded condition of door SZ1000516 and failed to implement a mitigating strategy. Corrective actions included restoration of the door to operable status and coaching of the responsible individual. Future corrective action will include implementation of a maintenance strategy for control room pressure boundary doors to ensure reliable operation. There were no nuclear safety consequences or radiological consequences as a result of this event. (4-2017) 366A U.S. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.qovireading-rm/doc-collectionsinureasistaff/sr1 0221r31) APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020 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 Infocoilects.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 DESCRIPTIONOn July 9, 2017, at approximately 2158 hours central daylight time a pressure boundary door to the Control Room Envelope (NA) was left unsecure. The Control Room Envelope was inoperable for approximately one minute, at which time the door was closed. Grand Gulf Nuclear Station personnel identified that a loss of Safety Function occurred due to a breach in the Control Room Envelope resulting in inoperability of both divisions of Standby Fresh Air systems. REPORTABILITYThis event is being reported under 10CFR50.73(a)(2)(v)(D), as any event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. CAUSEThe direct cause of the event was that the door was degraded, resulting in the door dragging on the floor such that the automatic door closer would not function properly. The organization failed to understand the nuclear safety consequence associated with the degraded condition and failed to implement a mitigating strategy. CORRECTIVE ACTIONSImmediate Corrective Actions:Upon identification the door was immediately secured and the Control Room Supervisor was notified. The event was entered into the Grand Gulf Nuclear Station corrective action program. Grand Gulf Nuclear Station Security personnel provided immediate coaching to the employee who left the door open. The support hinge bushing was replaced, restoring normal door closure. Corrective Actions To Prevent Recurrence: Implement a maintenance strategy for doors SZ1000715, Passageway Door; SZ1000708, Corridor Door; SZ1000612, Corridor No. 3 Door; SZ1000516, Corridor Door. NRC FORM(6-2016) nec.,„..., 366A U.S. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET(See NUREG-1022, R.3 for instruction and guidance for completing this form htlp://www.nrc.novireading-rm/doc-collections/nurecis/staff/sr1022/r3/) APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020
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 SAFETY SIGNIFICANCEThere were no nuclear safety consequences or radiological consequences as a result of this event. No Technical Specification Safety Limits were violated. PREVIOUS SIMILAR OCCURRENCESThe identified licensee event reports were reviewed and it has been determined that the causes and corrective actions for the previously identified events were sufficiently different that they could not have predicted or prevented the occurrence of this event. | |
05000440/LER-2017-002 | 27 June 2017 | Perry | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Reactor Protection System Main Turbine Reactor Recirculation Pump Control Rod | On April 30, 2017, at 1818 hours, while the plant was at 100 percent rated thermal power, main turbine steam bypass valve number 1 partially opened. Power was subsequently lowered in an attempt to close the bypass valve. While lowering power the bypass valve would shut and then reopen and power would again be lowered. When power was lowered to approximately 74 percent the bypass valve remained closed. During the transient the reactor protection system (RPS) trip functions for the main turbine stop valve closure and turbine control valve fast closure scram were declared inoperable due to the opening of the bypass valve, which changes the bypass setpoint for those RPS trips. With the loss of RPS trip capability. a loss of safety function existed intermittently for approximately 37 minutes. The manual reactor trip function and other RPS functions remained operable. Both channels of the rod withdrawal limiter (RWL) and the end of cycle reactor recirculation pump trip (EOC-RPT) function were also declared inoperable. These functions are credited in accident analysis and this also resulted in a loss of safety function in accordance with the plants Technical Specification bases. The direct cause of the bypass valve opening was degradation of the Primary Low Value Gate (PLVG) card in the main turbine speed control circuit. The safety significance of this event is considered to be small. This event is not considered a safety system functional failure as the specific functions were maintained and never bypassed during the event. This event is being reported under 50.73(a)(2)(v)(A) and 50.73(a)(2)(v)(D) for a loss of safety function. | |
05000440/LER-2017-003 | 26 June 2017 | Perry | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function | Primary containment Shield Building | On April 27. 2017, at 0545 hours, with the plant in Mode 1 at 100 percent rated thermal power. while shifting Annulus Exhaust Gas Treatment System (AEGTS) from sub-system B to sub-system A, annulus differential pressure could not be maintained within the required system operating band, which caused an unplanned entry into technical specification limiting conditions of operation and a momentary loss of safety function. Sub-system B was inoperable while being shutdown to standby readiness in accordance with plant operating procedures and would not have automatically started in response to an actuation signal. Investigation determined that sub- system A was inoperable due to failed recirculation damper. Sub-system B was restarted at 0550 hours. to maintain annulus differential pressure. Annulus differential pressure was maintained above the minimum differential pressure throughout the event and the technical specification limiting conditions for operation were not exceeded. The cause for the AEGTS sub-system A recirculation damper failure was a failure to follow procedure which resulted in the split coupling that connects the actuator to the damper not being properly tightened. The recirculation damper was repaired and AEGTS sub-system A was restored to operable status on April 28, 2017, at 2208 hours. The failure to follow procedure was addressed by the FENOC performance management process. Since AEGTS is not a core damage mitigation system and does not mitigate large and early containment releases, the inoperability of the AEGTS is determined to be of small safety significance. This event is being reported in accordance with 10CFR50.73(a)(2)(v)(C) and 10CFR50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfilment of a safety function. | |
05000528/LER-2017-001 | 11 April 2017 14 June 2017 | 14 June 2017 | Palo Verde Palo Verde Nuclear Generating Station Unit 1 | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Auxiliary Feedwater Containment Spray | On April 17. 2017, the staff identified a low refrigerant level in the Unit 1 train B essential chilled water (EC) system chiller during inspection. Operations personnel immediately declared EC chiller train B inoperable. On April 17, 2017, the leak was corrected and EC chiller train B was refilled with refrigerant to within the manufacturer's specifications. Operations personnel declared the system operable on April 18, 2017. The chiller had been inoperable since April 11. 2017, when the automatic purge system was placed into service. The direct cause of the low refrigerant level was leakage due to prior installation of a fitting on the automatic purge system filter without a plug. During the 7-day period that EC chiller train B was inoperable, the supported low pressure safety injection (LPSI) system train B was inoperable. LPSI train A was also inoperable for approximately 17 minutes on April 13, 2017, during the performance of a routine surveillance test. This 17-minute period represented a condition that could have prevented the fulfillment of a safety function. The cause of the leak was determined to be ineffective work instructions that did not identify the appropriate part number to be used during filter replacement. Corrective actions include revision of the work instructions. This change will ensure that the existing plug remains in place during filter element replacement. A leak test was also added to the work instructions to verify that no refrigerant leaks are present following maintenance. |
05000324/LER-2017-002 | 13 April 2017 12 June 2017 | 9 June 2017 | Brunswick | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | Emergency Diesel Generator Primary containment Residual Heat Removal | 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. |