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On December 9, 2017 at 1347 CDT the Main Control Room received annunciators that indicated a trip of a 4160V 1A1 Breaker, the 480V transformer 1A and Al feed breaker. The loss of Division 1 480V power caused the instrument air (IA) containment isolation valves to fail close as designed. The loss of IA affected various containment loads, including the scram pilot air header and containment isolation valves. Another consequence of this event was that secondary containment differential pressure became positive due to fuel building ventilation dampers failing closed by design due to the loss of power. Operations entered Emergency Operating Procedure (EOP) -08, Secondary Containment Control, and Technical Specification (TS) Limiting Condition for Operation (LCO), 3.6.4.1 Action A.1. Division 2 Standby Gas Treatment System was activated at 1350 and restored secondary containment differential pressure within allowable TS values at 1351. The TS LCO and EOP were exited when allowable TS values were restored. Due to the loss of IA, a manual reactor scram was inserted at 1353 when two control rods began drifting in as expected.

A phase to ground fault was identified on 480V transformer 1A (1AP11E). On December 14, the 480V transformer was replaced and the plant returned to Mode 1 operations on December 15. The condition described in this report was determined to be reportable under 10 CFR50.73(a)(2)(iv)(A), 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73 (a)(2)(ii)(B). The cause of the transformer failure is currently under investigation and will be provided in a supplemental report. This event is classified as an unplanned scram with complications due to the loss of the Division 1 480V power.

On November 5, 2017, at approximately 1240 CDT, the Main Control Room (MCR) received numerous annunciators that indicated a trip of the Emergency Reserve Auxiliary Transformer (ERAT) Static VAR Compensator (SVC) caused by a voltage transient on the 138 kV offsite supply. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.8.1, AC Sources-Operating, Required Action A.1 and A.2 were entered. As a result of the voltage transient, the Division 1 Fuel Building ventilation (VF) system isolation dampers closed causing a trip of VF supply and exhaust fans. With no operating VF fans, Secondary Containment (SC) vacuum rose to slightly greater than 0 inches water gauge (WG) at 1241 CDT which exceeded the TS requirement of greater than 0.25 inches vacuum WG. The MCR entered Emergency Operating Procedure (EOP)-8, Secondary Containment Control and TS LCO 3.6.4.1, Secondary Containment, Required Action A.1. The cause of the SC differential pressure becoming positive is that the circuit design of VF is not adequately robust to withstand loss of the 138 kV feed. At the time, the Division 1 safety bus was being fed from the ERAT.

Secondary Containment vacuum was restored within TS requirements at 1242 CDT by starting the Standby Gas Treatment System. A modification will be installed to prevent tripping VF during a momentary loss of power. Installation of a 138 kV Ring Bus is scheduled that is intended to improve the reliability of the radial feed of the 138 kV line. This event is being reported as a condition that could have prevented fulfillment of a safety function under 10 CFR 50.73(a)(2)(v)(C).

On September 27, 2017, at approximately 10:00 a.m. CDT, with the plant operating at 85 percent power, a door in the auxiliary building pressure boundary was left unsecured by an employee entering the building. The employee failed to fully close the door, and then did not properly challenge the door to confirm its security prior to leaving the area. A security officer responded to the resulting alarm, and fully closed the door approximately four minutes later. Since the worker had sufficient experience with watertight doors to know their proper operation, this event is considered a skill-based error caused by over-confidence and improper assumptions. Having successfully used such doors numerous times, the worker was confident in the ability to do so. The effort to check the door's security by pushing it failed, likely due to its heavy mass.

A briefing memorandum was issued from the general manager to site personnel. Tamper alarms were installed on all watertight doors in the secondary containment boundary to provide an audible indication that the door is open.

Administrative controls have been instituted to schedule routine battery replacements.

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)

On June 20, 2017, at 1444 hours (EDT), with the reactor at 100% core thermal power and steady state conditions, plant personnel notified the Main Control Room that both doors in the secondary containment airlock at the 23 foot elevation on the East Side Reactor Building (RB) Entrance were opened simultaneously.

The failure of this interlock (to prevent both doors from being opened) caused a loss of secondary containment per Technical Specification (TS) 3.7.C.1. The doors were immediately closed, and the secondary containment boundary was reestablished.

An 8-hour non-emergency notification was made in accordance with 10 CFR 50.72(b)(3)(v)(C), 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 control the release of radioactive material. The safety significance of this event was minimal given the impact on the secondary containment.

Secondary containment remained available and functional during the event since secondary containment was immediately restored by closing the doors.

On April 5, 2017, at 0030 hours (EDT) with the Reactor in the Run Mode at approximately 97 percent power, both trains of the Standby Gas Treatment System (SBGTS) were made inoperable during the performance of a surveillance test of secondary containment prior to the refueling outage, With both trains of SBGTS inoperable while in the Run mode, this event is reportable per the requirements of Title 10, Code of Federal Regulations (CFR) 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D), any event that could have prevented the fulfillment of the safety functions to "control the release of radioactive material" and "mitigate the consequences of an accident.

This event had no impact on the health and/or safety of the public.

On February 16, 2017, at 1320 EST with Unit 2 at 0 percent rated thermal power due to being in a refueling outage, maintenance electricians were sent to the field to perform a protective relay trip test for the 2D start-up transformer (SAT). During the test setup, the 2E 4160 VAC Emergency Bus was inadvertently and momentarily de-energized, causing the 2A Emergency Diesel Generator (EDG) to autostart, secondary containment to isolate, and start of the standby gas treatment system. Subsequent investigations revealed that the cause of the event was due to a movement operated contact (MOC) switch adapter was not required to be installed on the 2D normal supply breaker in the 2E 4160 VAC bus. All systems responded appropriately.

A review of the event determined that the MOC switch adapter was not required to be installed by the procedure, but was instructed to be installed by supervision. Corrective actions were taken to cover supervisor roles and responsibilities and the need for all workers to follow plant standards for procedure use and adherence. All breaker procedures and protective relay test procedures were reviewed to determine if a MOC switch adapter needs to be installed. Continuing training will also be held to cover this event and its lessons learned.

On March 24, 2017, it was determined that the A Standby Gas Treatment System (SGTS A) was not operating as expected. The investigation into the event revealed the Single Nest Power Supply had failed resulting in loss of flow control. It was determined the power supply had been replaced and the technical specification limiting condition for operation exited on February 23, 2017. The A SGTS was not run between replacement of the power supply and the time of discovery condition on March 24, 2017. Additionally, the B SGTS was removed from service for planned corrective and preventative maintenance on February 28, 2017 and returned to service on March 3, 2017. This condition prohibited by technical specifications in accordance with 10 CFR 50.73(a)(2)(i)(B) for the A SGTS train being inoperable for a period great than allowed by technical specifications. The cause of this event has been determined to be a power supply that could not be fully inserted due to pre-existing damage.

The damaged power supply was not appropriately corrected prior to installation due to an incorrect screening practice. The defective power supply was replaced and tested satisfactorily. Entergy established proceduralized barriers to minimize recurrence of similar errors through the establishment of pre-installation checks for the power supply as well as post maintenance testing of the replaced power supplies. There were no actual nuclear safety consequences or radiological consequences. No Technical Specification Safety Limits were violated.

(4-2017) to 366A U.S. NUCLEAR REGULATORY COMMISSION

CONTINUATION SHEET

(See NUREG-1022, R.3 for instruction and guidance for completing this form htiorAww.nrc.00vireadino-rrradoc-collectionsinureosistaff/sr1022,ra) 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 20555-8001, or by e-mail to Infocollects.Resourcee 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.

Grand Gulf Nuclear Station, Unit 1 05000 416

DESCRIPTION

On March 24, 2017, while operating the A Standby Gas Treatment System (SGTS A) the SGTS Flow Recorder was indicated a downscale reading. The Plant Data System computer points indicated a flow mismatch between the available indications of approximately 5000 cubic feet per minute (CFM). Due to the mismatch indications and the downscale reading the surveillance was terminated and Standby Gas "A" was returned to standby.

The investigation of the identified conditions determined the failure of the systems was the Single Nest Power Supply. Failure of this power supply would result in a loss of the associated train's flow control. The investigation revealed that the power supply was replaced and the technical specification limiting condition for operation exited on February 23, 2017. The A SGTS was not run between replacement of the power supply and the time of discovery condition on March 24, 2017.

Prior to the power supply being replaced the system had been successfully tested and therefore it was determined that this condition could only have been present since the installation of the new Power Supply was completed on February 23, 2017.

The investigation further revealed that the B SGTS was removed from service for planned corrective and preventative maintenance on February 28, 2017 and returned to service on March 3, 2017.

The above described condition rendered the A SGTS inoperable and also resulted in a period when both SGTSs were inoperable during the same time period.

REPORTABILITY

The condition is also reportable as a condition prohibited by technical specifications in accordance with 10 CFR 50.73(a)(2)(i)(B) for the A SGTS train being inoperable for a period great than allowed by technical specifications.

An engineering evaluation was performed that demonstrated the SGTS A was able to perform its safety function with the identified nonconformance. Therefore this LER supplement retracts reporting this concern as a loss of safety function under 10 CFR 50.73(a)(2)(v)(C).

CAUSE

The cause of this event has been determined to be within the control system of the Single Nest Power Supply.

The equipment failure analysis determined the power supply could not be fully inserted due to pre-existing damage. The damage to the power supply was not appropriately corrected prior to installation due to an incorrect screening of the condition report that initially identified the pre-existing condition.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-2016) FiEnt, LICENSEE EVENT REPORT (LER) (A 2 CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form http:ilwwwnrc.govireadina-rinidcc-collectionsinureasistaffisrl 022/0) 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 Infacollects.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.

Grand Gulf Nuclear Station, Unit 1 05000 416

3. LER NUMBER

CORRECTIVE ACTIONS

The defective power supply was replaced and tested satisfactory.

Proceduralized barriers were established to minimize recurrence of similar errors through the establishment of pre-installation checks for the power supply as well as post maintenance testing of the replaced power supplies.

The governance was reviewed as it relates to discovery of deficient conditions on replacement parts and required actions based on classification of the part (i.e. safety related, quality part, etc.). This action was performed to determine if this issue should have been identified as a NON CONFORMING part and if it should have been tagged and segregated. This review revealed the process was not followed and follow-up actions were developed to correct the cause of the error.

SAFETY SIGNIFICANCE

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

PREVIOUSLY SIMILAR EVENTS

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 the morning of June 7, 2016, while operating at 100% power, workers opened doors concurrently when entering a secondary containment access airlock. The individuals involved each closed their respective doors upon encountering this unexpected condition; however, the result was a brief inoperability of secondary containment.

This resulted in an 8 hour reportable event. The Resident Inspector was notified, and an Event Notification was made pursuant to 10 CFR 50.72(b)(3)(v)(C) due to a condition at the time of discovery that prevented the fulfillment of the Secondary Containment safety function (Reference ENS 51985). Following the event, the doors functioned properly, and no deficiencies were noted with either door.

There were no radiological releases associated with this event.

On January 29, 2016, at 1518 CST, with the plant in cold shutdown, power was lost on reserve station service (RSS) line no. 1. This is one of two sources of offsite power required by Technical Specifications. The power loss de-energized the Division 1 onsite AC safety.- related switchgear, causing an automatic start of the Division 1 emergency diesel generator (EDG). The Division 1 reactor protection system (RPS) bus was also de-energized, causing a half-scram signal. Approximately 8 minutes later, a full actuation of the RPS occurred due to high water level in the control rod drive hydraulic system scram discharge volume header. All reactor control rods were already fully inserted. The loss of Division 1 RPS also caused the actuation of the Division 1 primary containment isolation logic. The Division 1 isolation valves in the balance-of-plant systems closed as designed. Both trains of the standby gas treatment system actuated.

The loss of RSS No. 1 was caused when company transmission department personnel working in the local 230kV switchyard executed a deficient work instruction while modifying relay settings. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as the automatic actuation of the Division 1 EDG, the Division 1 primary containment logic, and the reactor protection system (while subcritical). At the time of the event, the shutdown cooling system was operating on the Division 2 subloop, which was unaffected.

The Division 1 EDG performed as designed. This event was, thus, of minimal significance to the health and safety of the public.

building. The alarms occurred during an entry to the Unit 2 Reactor Water Cleanup (RWCU) pump room. A negative reactor building pressure was restored within two minutes (approximately 20:40 hours) without operator action.

Since both Units 1 and 2 share a common reactor building (RB), the loss of differential pressure impacted both Units 1 and 2 secondary containments.

The cause was a sheared air line inside the Unit 1 RB ventilation exhaust plenum which depressurized the air header supplying operating air to all three Unit 1 reactor building exhaust fan isolation dampers, causing the dampers to fail open, including the one on the standby fan.

Corrective actions included replacing the sheared air line, and the addition of a preventive maintenance task for replacement of equivalent air lines on all RB supply and exhaust fan dampers.

The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10CFR 50.73(a)(2)(v)(C), 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 control the release of radioactive material.

On September 22, 2015 at 17:03, with James A. FitzPatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated a spike in Secondary Containment (SC) differential pressure (d/P) during performance of a surveillance test associated with automatic isolation of SC and initiation of the Standby Gas Treatment System. Per the plant data systems SC d/P exceeded the Technical Specification (TS) allowed value, and then immediately trended negative following auto-start of one of the trains of Standby Gas Treatment.

The time period that SC d/P was greater than the TS allowed value is reportable pursuant to 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.73(a)(2)(v)(C), as an event or condition that could have prevented fulfillment of a safety function. SC was operable following reestablishment of greater than or equal to 0.25 inches of water vacuum, and remains operable.

SC d/P excursions during transition from normal to isolation mode of the Reactor Building Ventilation (RBV) System are an expected condition, and attributable to the design of the non-safety related RBV System. The cause of the SC d/P exceeding the TS allowed value has been determined not to be associated with a component failure or equipment malfunction. Similar reportable events were identified during preparation of this report. A comprehensive listing of these occurrences is included in the report.