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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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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

DESCRIPTION

On 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.

REPORTABILITY

This 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.

CAUSE

The 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 ACTIONS

Immediate 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

• Reported :essons 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 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.

2. DOCKET 3. LER NUMBER 05000 416

SAFETY SIGNIFICANCE

There were no nuclear safety consequences or radiological consequences as a result of this event. No Technical Specification Safety Limits were violated.

PREVIOUS SIMILAR OCCURRENCES

The 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.

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.

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.