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 Entered dateSiteRegionReactor typeEvent description
ENS 5339912 May 2018 06:58:00Grand GulfNRC Region 4GE-6On 5/11/2018, at 2327 hours CDT, with the plant in Mode 5, Grand Gulf Nuclear Station was making preparations for surveillance test 06-OP-1P75-R-0003, Standby Diesel Generator 1 Functional Test. The Grand Gulf Nuclear Station experienced an auto-start of the Division 1 (Emergency) Diesel Generator (EDG) when the 15AA Bus Potential Transformer (PT) fuse drawer was racked out instead of the line PT fuse drawer for Bus 15AA feeder breaker 152-1514. This resulted in the 15AA Incoming Feeder Breaker 152-1511 from Engineered Safety Features Transformer 12 opening, de-energizing the 15AA Bus. The Division 1 EDG started and energized Bus 15AA. The Division 1 LSS SYSTEM FAIL annunciator was received and Standby Service Water A failed to start due to the 15AA Bus PT fuse drawer being racked out. Standby Gas Treatment Train B was manually initiated per the Loss Of AC Power Off Normal Emergency Procedure. Station equipment operated as expected based on the PT fuse drawer that was racked out. The Division 1 EDG was manually tripped from the Control Room because cooling from the Standby Service Water A was not available. RHR (residual heat removal) B was in Shutdown Cooling (mode) and was verified not affected The licensee has notified the NRC Resident Inspector.
ENS 533824 May 2018 13:50:00River BendNRC Region 4GE-6During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, River Bend Station identified non-conforming conditions in the plant design such that specific TS equipment is considered to not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Diesel Generator Building through conduit and pipe penetrations. A tornado could generate multiple missiles capable of striking Division 1, Division 2, and Division 3 Diesel Generator support equipment rendering all Safety Related Diesel Generators inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for 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 (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector.
ENS 533741 May 2018 20:42:00Grand GulfNRC Region 4GE-6At 1551 hrs (CDT) on 5/1/2018, with the plant in Mode 5, a division one Reactor Pressure Vessel (RPV) Level 1 signal was received; however there was no actual change in RPV level. RPV Level remained at High Water Level supporting refuel operations. This caused an actuation of division one Load Shed and Sequencing system that shed and then re-energized the 15 bus. Division one diesel generator started from standby. Residual Heat Removal pump 'A', which was in shutdown cooling mode, was lost during the bus shed, and was re-sequenced upon re-energization of the 15 bus. Upon restoration of shutdown cooling, the RHR pump discharged into the RPV. RCS temperature increased approximately 5 degrees Fahrenheit as a result of the loss of shutdown cooling. The cause of the actuation signal is under investigation. In accordance with NUREG 1022, Event Reporting Guidelines, this event is conservatively reported under 10 CFR 50.72(b)(2)(iv)(A) as an event that results in emergency core cooling system discharge into the RCS as a result of a valid signal, under 10 CFR 50.72(b)(3)(iv)(B)(8) as an event that results in the actuation of emergency ac electrical power systems, and under 10 CFR 50.72(b)(3)(v)(B) as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function (remove residual heat). The licensee notified the NRC Resident Inspector.
ENS 5336526 April 2018 18:50:00River BendNRC Region 4GE-6River Bend Station experienced an inadvertent initiation and injection of High Pressure Core Spray (HPCS) at 1531 (CDT) on 4/26/2018 while operating at 100 percent power. During replacement of Level Transmitter B21-LTN081C 'Reactor Vessel Low Water Level 1', Main Control Room received an inadvertent initiation and injection of High Pressure Core Spray. The HPCS injection valve was open for approximately 40 seconds before the operators manually closed the valve. Feedwater Level Control responded per design and maintained Reactor Water Level nominal values. The Division 3 Diesel Generator (DG) also automatically started in response to the actuation signal. The DG did not automatically connect to the Division 3 switchgear since there was not a low voltage condition on the bus. The manual closure of the injection isolation valve caused the system to be incapable of responding to an automatic actuation signal. The manual override of the injection isolation valve was reset approximately 16 minutes after the event, restoring the system to its standby condition. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a condition that caused ECCS (Emergency Core Cooling System) discharge to RCS (Reactor Coolant System) and 10 CFR 50.72(b)(3)(v)(D) as a condition that caused the loss of function of the HPCS System. The Senior NRC Resident inspector has been notified.
ENS 5333513 April 2018 21:04:00Grand GulfNRC Region 4GE-6

At 1208 CDT on April 13, 2018, GGNS (Grand Gulf Nuclear Station) identified cracks in the primary containment concrete penetration (outer wall) around feed water line 'B'. There are no available dimensions for crack width or depth until further inspections are performed. In accordance with NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73, Section 3.2.4, any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded, requires that when a principal safety barrier is declared inoperable the condition must be reported under 10 CFR 50.72 (b)(3)(ii). The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM GERRY ELLIS TO HOWIE CROUCH AT 2012 EDT ON 4/15/18 * * *

Grand Gulf Nuclear Station (GGNS) personnel performed an inspection of the wall around feed water line 'B'. This inspection included the protective coating in the identified area and a partial inspection of the underlying concrete. The inspection of the protective coating found a collection of non-linear anomalies, chipping, and flaking. The inspection found non-significant linear indications in the concrete. Grand Gulf Nuclear Station determined that the collection of non-significant coating imperfections and non-significant indications in the concrete do not constitute serious degradation of primary containment. The indications do not adversely impact the operability, mission time, or safety-function (as described per Technical Specification 3.6.1.1, Primary Containment) of the containment structure. The as-found conditions have been entered into the GGNS corrective action program for final disposition. The containment structure is operable, therefore, GGNS is retracting this event notification. The licensee has notified the NRC Resident Inspector. Notified R4DO (Kellar).

ENS 5332411 April 2018 10:14:00River BendNRC Region 4GE-6At time 0150 CDT on April 11, 2018, a condition was identified that could impair the ability of the Control Building Air Conditioning System to perform its design function. Engineering determined that the time delay relays HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) could fail in a manner that challenges the design safety function of the Control Building Chilled Water System during a Loss of Offsite Power (LOP) Event. A failure of the time delay relay HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) to provide the time delay function would cause both the Division I and Division II HVK chilled water pumps to start after a LOP, which in turn could hinder the auto start of either Division I or Division II chillers. Currently the Chilled Water System is otherwise operating as designed. All operator actions are in place to ensure the plant meets all required designed safety system functions. Work is currently underway to correct this design vulnerability. The NRC Resident Inspector has been notified of this condition.
ENS 533175 April 2018 18:23:00Grand GulfNRC Region 4GE-6On Thursday, April 5, 2018, at approximately 1117 hours Central Daylight Time, Entergy contract personnel opened the personnel hatch allowing access to the roof of the Secondary Containment Building for the purposes of performing an inspection of various items located on the roof. During the time period the individuals were on the roof, the hatch was left open. An individual was adjacent to the door with a radio and had constant communication link with the control room operator. Pursuant 10 CFR 50.72(b)(3)(v)(C), and 10 CFR 50.72(b)(3)(v)(D) this event is being reported as an event or condition that could have prevented the fulfillment of a safety function. Because the site had an individual briefed and at the door in constant communications with the control room to close the hatch if condition required such an action, this event is not viewed as an actual loss of safety function. The NRC Resident Inspector was notified.
ENS 5330431 March 2018 10:00:00Grand GulfNRC Region 4GE-6At 0206 (CDT) on March 31, 2018, with the plant in Mode 1 at 100% rated core thermal power, Grand Gulf Nuclear Station experienced a loss of Secondary Containment. During the performance of a Standby Gas Treatment System (SGTS) drawn down test with Auxiliary Building train bay door (1A319A) as the secondary containment boundary, Grand Gulf was unable to maintain secondary containment pressure, as required by SR (surveillance requirement) 3.6.4.1.4, greater than or equal to 0.266 inches of water vacuum for 1 hour. Following initial vacuum draw down, secondary containment pressure degraded to 0.225 inches of water vacuum with operators in the field reporting air leakage from door 1A319A. The test was secured and Secondary Containment was declared inoperable and Technical Specification 3.6.1.4 A.1 was entered. Following completion of the failed surveillance test, Secondary Containment was returned to an operable status at 0315 hours on March 31, 2018, by returning the system to a previously known operable configuration by closing doors 1A310, 1A312 and 1A319. This is being report under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector.
ENS 5330330 March 2018 20:47:00ClintonNRC Region 3GE-6On March 30, 2018 at 1305 CDT, with the reactor at 98 percent core thermal power and steady state conditions, plant personnel identified that both doors of the containment personnel airlock were open simultaneously due to failure of the interlock. Personnel were at both the outside and inside doors. Immediate action was taken to close the inner containment personnel airlock door and it was verified closed. Both doors of the containment personnel airlock were open for less than one minute. There was no radioactive release as a result of the event. The cause of the interlock failure is under investigation. This condition requires an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(ii)(A), the condition of the nuclear power plant, including its principal safety barriers (primary containment), being seriously degraded. This condition is also reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The NRC Resident Inspector was notified.
ENS 5322723 February 2018 18:04:00Grand GulfNRC Region 4GE-6On February 18, 2018, Grand Gulf Nuclear Station experienced the concurrent inoperability of two Emergency Diesel Generators (DG). This event is being reported as a late 8-hour non-emergency notification per 10 CFR 50.72(b)(3)(v)(D) as an 'Event or Condition that Could Have Prevented Fulfillment of a Safety Function (Accident Mitigation).' On February 14, 2018 at 0100 (CST), the Division 2 Diesel Generator was declared inoperable, and subsequently removed from service for maintenance. On February 18, 2018 at 0006 (CST), the Division 3 Diesel Generator Jacket Water temperature exceeded the trip setpoint and Division 3 Diesel Generator was declared inoperable. The Division 2 Diesel Generator was restored and declared operable on February 18, 2018 at 0355 (CST), and the Division 3 Diesel Generator was restored and declared operable on February 18, 2018 at 1240 (CST). As a result, Technical Specification Condition 3.8.1.E was entered at 0006 (CST) on February 18, 2018 and exited at 0355 (CST) on February 18, 2018. Technical Specification Bases 3.8.1.E.1 states 'With two DGs inoperable, there is one remaining standby AC source. Thus, with an assumed loss of offsite electrical power, insufficient standby AC sources are available to power the minimum required ESF functions.' Offsite power was available throughout this event and there was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5320110 February 2018 22:37:00Grand GulfNRC Region 4GE-6On 2/10/18 at 1835 CST at Grand Gulf Nuclear Station, while the 208 ft. Containment Airlock Outer Door was tagged-out for planned maintenance, the 208 ft. Containment Inner Door was determined to be inoperable. Grand Gulf had performed 06-ME-1M23-R-0001, Personnel Airlock Door Seal Air System Leak Test, on the 208 ft. Containment Airlock Inner Door which had been deemed satisfactory. While performing planned maintenance on the outer door an additional review of the paperwork determined that the test was actually unsatisfactory on the inner door. TS 3.6.1.2 Condition C was entered at 1835 CST on 2/10/18 for both 208 ft. Containment Airlock Doors being inoperable. Maintenance of the Outer Door is expected to be completed, and the airlock returned to operable status, prior to TS required action completion time. The licensee notified the NRC Resident Inspector.
ENS 531921 February 2018 14:23:00River BendNRC Region 4GE-6

At 1057 CST on February 1, 2018 with the unit in Mode 1 at approximately 27% power, a manual actuation of the Reactor Protection System (RPS) was initiated due to an unexpected trip of the B Recirc Pump with A Recirc Pump in fast speed. B Recirc Pump tripped during transfer from slow to fast speed resulting in single loop operation. Operators were unable to reconcile differing indications of core flow. This resulted in a conservative decision to initiate a manual scram. The cause of the B Recirc Pump trip and the apparent issues with core flow indication are under investigation. The plant is currently stable in Mode 3. The plant response to the scram was as expected. All control rods (fully) inserted as expected; the feedwater system is maintaining reactor vessel water level in the normal control band and reactor pressure is being maintained with steam line drains and main turbine bypass valves. The NRC Senior Resident (Inspector) has been notified.

  • * * RETRACTION AT 1015 EDT ON 03/22/2018 FROM DAVID DABADIE TO OSSY FONT * * *

This event was initially reported under 10 CFR 72(b)(2)(iv)(B) as a manual actuation of the RPS due to an unexpected trip of the B Reactor Recirculation Pump with the A Reactor Recirculation Pump running in fast speed (Single Loop Operations). Operations was unable to reconcile differing indications of core flow and made the conservative decision to perform a planned shutdown in accordance with normal operating procedures. Therefore, this event 'resulted from and was part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.73(a)(2)(iv)(A) and NUREG-1022 Section 3.2.6. Consequently, this event is not reportable as an actuation of RPS. The NRC Resident Inspector has been notified. R4DO (Groom) has been notified.

ENS 5318830 January 2018 21:56:00Grand GulfNRC Region 4GE-6On 1/30/2018 at 1750 (CST), the Reactor Pressure Control Malfunctions ONEP (Off Normal Event Procedure) was entered due to main turbine load oscillations of approximately 30 MWe peak to peak. At 1822 (CST), a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown due to continued main turbine load oscillations. Reactor SCRAM ONEP, Turbine Trip ONEP, and EP-2 were entered. Reactor water level was stabilized at 36 inches narrow range on startup level and reactor pressure stabilized at 933 psig using main turbine bypass valves. Reactor Water Level 3 (11.4 inches) was reached which is the setpoint for Group 2 (RHR to Radwaste Isolation) and Group 3 (Shutdown Cooling Isolation). No valve isolated in these systems due to all isolation valves in these groups being in their normally closed position. The lowest Reactor Water level reached was -36 inches wide range. No other safety system actuations occurred and all systems performed as designed. That event is being reported under 10CFR 50.72(b)(2)(iv)(B) as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical and also reported under 10CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of RPS. The MSIVs are open with decay heat being removed via steam to the main condenser using the bypass valves. Off site power is stable, and the plant is in a normal shutdown electrical lineup. RCIC (Reactor Core Isolation Cooling) was out of service for maintenance, and the reactor water level did not reach the system activation level. The cause of the main turbine load oscillations being investigated. The licensee notified the NRC Resident Inspector.
ENS 5313723 December 2017 05:00:00PerryNRC Region 3GE-6High Pressure Core Spray System was declared inoperable due to the discovery of a through-wall leak on the Minimum Flow line. Leak rate is 60 drops per minute from ASME Class 2 Piping. The leak has been isolated and the High Pressure (Core Spray) System has been placed in Secured Status. High Pressure Core Spray is considered a single train safety system. Inoperability of (the) High Pressure Core Spray System is considered an 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. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Senior Resident Inspector was notified. Technical Specifications Limiting Condition for Operation 3.5.1 Condition B was entered, requiring restoration of the High Pressure Core Spray System in 14 days. The licensee plans to notify State and Local Governments (Lake, Geauga, and Ashtabula Counties).
ENS 5311712 December 2017 20:34:00Grand GulfNRC Region 4GE-6At approximately 1330 CST on Tuesday, December 12, 2017, Grand Gulf Nuclear Station declared Division 3 'C' Battery inoperable due to questions concerning battery terminal connection continuity. Technical Specification 3.8.4, DC Sources - Operating, Condition E, Required Action E.1, requires the station to declare the High Pressure Core Spray System inoperable immediately. The Division 3 'C' Battery and High Pressure Core Spray System was declared operable and the LCOs (Limiting condition of operation) were declared met at 1731CST on Tuesday, December 12, 2017. Based on field measurements of terminal torque and resistance, the as-found and as-left terminal resistance micro-ohm readings indicated satisfactorily all times. Formal evaluation of the as-found condition of the battery is in progress. This report is to notify the NRC of a loss of safety function on the High Pressure Core Spray System. The NRC Resident Inspector was notified.
ENS 5311512 December 2017 17:40:00Grand GulfNRC Region 4GE-6At approximately 0918 CST on Tuesday, December 12, 2017, the Grand Gulf Nuclear Station experienced a loss of the Engineered Safety Features (ESF) Transformer 11 which was powering the Division 1 ESF bus. Subsequently, the station experienced an automatic start of the Division 1 Emergency Diesel Generator (EDG), partial isolation of the primary and secondary containment buildings and the isolation of the Reactor Core Isolation Cooling System (RCIC). It is not currently understood why the RCIC system isolated during this event. A team is investigating this issue separately from the loss of the ESF 11 transformer. The cause of the event is under investigation at this time. No other issues or unexpected events occurred. The NRC Resident Inspector has been notified of the event.
ENS 531109 December 2017 18:42:00ClintonNRC Region 3GE-6

At approximately 1347 (CST) on 12/09/17, the Main Control Room received annunciators that indicated a trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1 breaker. Numerous Division 1 components lost power (powered from unit subs 1A and A1). The Division 1 containment Instrument Air isolation valves had failed closed by design due to the loss of power. Due to the loss of containment instrument air, several control rods began to drift into the core as expected and, by procedure, the reactor mode switch was placed in the shutdown position at 1353 (CST). All control rods fully inserted. Also due to the loss of power, the Fuel Building ventilation dampers failed closed by design. With the normal ventilation system secured, secondary containment differential pressure rose to slightly greater than 0 inches water gauge which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge at 1348 (CST). The Control Room entered EOP-8, Secondary Containment Control. Secondary Containment differential pressure was restored within Technical Specification requirements at 1351 (CST) by starting the Division 2 Standby Gas Treatment system. This event is being reported as a manual actuation of the Reactor Protection System (RPS) and as a Condition that Could Have Prevented Fulfillment of a Safety Function.

The cause is currently under investigation. The NRC Resident has been notified. The licensee informed the NRC Resident Inspector.

  • * * UPDATE FROM DALE SHELTON TO VINCE KLCO AT 1658 EST ON 12/10/2017 * * *

During a review of plant logs it was identified that the primary to secondary containment differential pressure was identified to be outside of Technical Specification 3.6.1.4 limits of 0 plus or minus 0.25 psid at 2009 on 12/9/17 due to the primary containment ventilation system dampers closing as a result of the loss of power. This parameter is an initial safety analysis assumption to ensure that primary containment pressures remain within the design values during a Loss of Coolant Accident (LOCA). As a result, this condition is reportable as an unanalyzed condition that significantly degrades plant safety. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).

  • * * UPDATE FROM MICHAEL ANTONELLI TO VINCE KLCO ON 12/11/17 AT 1805 EST * * *

During the post transient review of the trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1, it was identified that the unplanned INOPERABILITY of the Low Pressure Core Spray (LPCS) system due to the loss of power to the injection valve constitutes an event or condition that could have prevented fulfillment of a safety function and is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. The High Pressure Core Spray (HPCS) remained available to perform the core spray function, if necessary, during a design basis Loss of Coolant Accident (LOCA), however HPCS and LPCS are each considered single train safety systems. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).

ENS 5309025 November 2017 06:02:00Grand GulfNRC Region 4GE-6

At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. At 0149 (CST), with reactor power just above the point of adding heat, IRM (Intermediate Range Monitor) channels A, C, and D received a spurious upscale trip signal which immediately cleared. Upon investigation, operability of RPS (Reactor Protection System) scram function for Intermediate Range Detectors was placed in question. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. The licensee notified the NRC Resident Inspector.

  • * * UPDATE ON NOVEMBER 26, 2017, AT 1850 FROM GRAND GULF TO MICHAEL BLOODGOOD * * *

At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. At 0149 (CST), with reactor power just above the point of adding heat, Intermediate Range Monitor neutron flux detector (IRM) channels A, C, and D received a spurious Upscale Trip signal which immediately cleared. Upon investigation, IRM channels A, C, and D were declared Inoperable. IRM G was already Inoperable for another reason. RPS scram function from IRM channels B, E, F, and H was always Operable and available. That event is being reported under 10CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. This Revised Statement to Event Notification # 53090 is being made to make it clear that only four IRM channels (A, C, D, G) were Inoperable and that the IRM RPS SCRAM function was still available from the four remaining Operable IRM channels (B, E, F, and H). The licensee notified the NRC Resident Inspector. Notified R4DO (O'Keefe)

  • * * RETRACTION ON 01/16/2018 AT 1629 EST FROM JASON COMFORT TO DAVID AIRD * * *

On 11/25/17, at 0149 (CST), with reactor power just above the point of adding heat, Intermediate Range Monitor neutron flux detector (IRM) channels A, C, and D received a spurious Upscale Trip signal which immediately cleared. Upon investigation, IRM channels A, C, and D were declared Inoperable. IRM G was already Inoperable for another reason. At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. RPS scram function from IRM channels B, E, F, and H was always Operable and available. That event was initially being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. After the trip alarms were received, the Operators spent approximately twenty minutes investigating possible causes and implications, and consulted with Reactor Engineering and the Shift Technical Advisor. The investigation showed that the plant was stable and the upscale IRM alarms were spurious. A review of plant technical specifications by the operators determined that a plant shutdown was not required. After further discussions, Operations concluded that a shutdown to allow further investigation of the issue was the prudent course of action. Prior to shutting down, Operations spent approximately twenty minutes reviewing procedures, notifying personnel to exit containment, and conducting a brief. The shutdown was then conducted by inserting a manual reactor scram by placing the reactor mode switch in SHUTDOWN. This was initially reported under 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the RPS. Based on the sequence of events, and Operator actions in conducting the shutdown, the event is considered 'part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.73(a)(2)(iv)(A). In accordance with NUREG-1022, Section 3.2.6, the event is not reportable as an actuation of RPS. The licensee notified the NRC Resident Inspector. Notified R4DO (Taylor).

ENS 5303826 October 2017 18:18:00Grand GulfNRC Region 4GE-6

At 1055 (CDT), drywell purge supply/initial vacuum relief 1E61F003B was declared INOPERABLE for Drywell Vacuum Relief System while performing a monthly surveillance. 1E61F003B is a Division II powered valve. Division 1 Emergency Diesel Generator is INOPERABLE due to a tagout. At 1455, under LCO 3.8.1.B.2 the station declared both divisions of Drywell Vacuum Relief INOPERABLE. GGNS (Grand Gulf Nuclear Station) identified that a loss of Safety Function occurred due to a loss of two 10-inch vacuum relief lines from the Drywell required by Technical Specification 3.6.5.6 and therefore could have prevented fulfillment of its safety function. That event is being reported under 10CFR50.72(b)(3)(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. Both divisions inoperable has placed the plant in a 72-hr. LCO shutdown action statement. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM RALPH FLICKINGER TO HOWIE CROUCH AT 1643 EST ON 12/15/17 * * *

At 1055 (CDT on 10/26/17), Drywell purge supply/initial vacuum relief 1E61F003B was declared INOPERABLE for Drywell Vacuum Relief System while performing a monthly surveillance. 1E61F003B is a Division II powered valve. Division 1 Emergency Diesel Generator was INOPERABLE due to a tagout. At 1455 (CDT on 12/26/17), under LCO 3.8.1.B.2, the station declared both divisions of Drywell Vacuum Relief INOPERABLE. GGNS identified that a loss of Safety Function occurred due to a loss of two 10-inch vacuum relief lines from the Drywell required by Technical Specification 3.6.5.6 and therefore could have prevented fulfillment of its safety function. This was initially reported under 10 CFR 50.72(b)(2)(v)(D). Div. 1 EDG was initially taken out of service at 1455 on 10/26 for preplanned maintenance (OP-EVAL). It was subsequently declared INOPERABLE-INOP due to a visible flaw indication in the exhaust manifold (1117 on 10/27). A subsequent Maintenance Functional Failure Evaluation and Past Operability Determination concluded the diesel was capable of performing its intended function for the required mission time, and therefore met the definition of OPERABLE. NUREG-1022 provides clarification for 10 CFR 50.72(b)(2)(v). NUREG Section 3.2.7, paragraph 4, states '...unless a condition is discovered that would have resulted in the system (otherwise) being declared inoperable, reports are not required when systems are declared inoperable solely as a result of Required Actions for which the bases is the assumption of an additional random single failure (i.e., . ..LCO 3.8.1, 'AC Sources Operating,' Required Actions .., B.2, or C.1). Per ACTION 3.8.1 .B.2, both trains of Drywell vacuum and Drywell Purge were inoperable for the purposes of Tech Specs. However, the normal power supply was available to Division 1 and there were no conditions which would have rendered the Division 1 diesel inoperable. Therefore, per Section 3.2.7 of NUREG-1022, this was not a Loss of Safety Function and was not reportable under 50.72(b)(2)(v)(D). The licensee has notified the NRC Resident Inspector. Notified R4DO (Deese).

ENS 530004 October 2017 05:53:00PerryNRC Region 3GE-6

On October 4, 2017, at 0250 hours (EDT), the Perry Nuclear Power Plant commenced a Technical Specification (TS) shutdown by lowering reactor power from 100 percent rated thermal power to 98 percent to comply with TS LCO 3.0.3. Reactor power was further reduced to 82 percent rated thermal power at 0430 hours (EDT). The plant had entered TS 3.0.3 at 0155 hours (EDT) upon loss of MCC (Motor Control Center), Switchgear, and Miscellaneous Electrical Equipment Areas HVAC System train A while train B was removed from service for maintenance. MCC switchgear ventilation train A was declared inoperable based on excessive belt noise and a dropped belt on MCC switchgear supply fan A. This also constitutes a loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(2)(i) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector was notified.

  • * * UPDATE ON 10/04/17 AT 0926 EDT FROM DAN HARTIGAN TO STEVEN VITTO * * *

Due to the loss of both trains of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC, actions were taken in LCO 3.8.7 for AC and DC Distribution Systems, LCO 3.8.4 for DC Sources, LCO 3.8.1 for AC Sources, and the associated support systems, the High Pressure Core Spray system was also declared inoperable, which is a single train safety system and therefore, an additional loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B), 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D). At 0620 hours (EDT) the A train of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC and High Pressure Core Spray was declared operable and LCO 3.0.3 was exited. The plant was restored to 100% (percent) power at 0804 (EDT). The NRC Resident Inspector was notified. Notified R3DO(Hills).

ENS 5299527 September 2017 14:26:00River BendNRC Region 4GE-6Security personnel reported to the Main Control Room that at time 1000 CDT (on 9/27/2017), an alarm indicated that a secondary containment door was open beyond the normal delay time allowed for entry and exit. Security personnel responded and found the door open and unattended with the dogs extended meaning that the door was unable to be closed. Security personnel secured the door at time 1004 CDT. No deficiencies were found with the door. The fact the door was open and unattended beyond the time allowed for normal entry and exit results in Technical Specification 3.6.4.1 'Secondary Containment-Operating,' not being met because surveillance requirement SR 3.6.4.1.3 is not met. This surveillance requires that doors be closed except during normal entry and exit. By definition in NUREG-1022, when Secondary Containment is inoperable, it is not capable of performing its specified safety function which in turn makes this condition reportable in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified.
ENS 5293629 August 2017 14:42:00Grand GulfNRC Region 4GE-6

On August 22, 2017 at 2321 hours, Grand Gulf Nuclear Station entered Technical Specification conditions for three Limiting Condition for Operations (LCOs) not met due to Residual Heat Removal 'A' (RHR 'A') being declared inoperable. LCOs not met:

  1) 3.5.1 for one low pressure ECCS (Emergency Core Cooling System) injection/spray subsystem.
  2) 3.6.1.7 for one RHR containment spray subsystem, and
  3) 3.6.2.3 for one RHR suppression pool cooling subsystem.

The station has made the decision to shutdown the plant based on the results of troubleshooting performed on the RHR 'A' pump. The restoration of RHR 'A' pump will not be completed prior to the end of the 7 day LCO completion time. Grand Gulf Nuclear Station initiated plant shutdown required by Technical Specifications 3.5.1, 3.6.1.7, and 3.6.2.3 at 1200 hours CDT on 08/29/2017 due to expected restoration of RHR 'A' exceeding the completion time of 7 days prior to restoring operability. The licensee notified the NRC Resident Inspector.

ENS 5292123 August 2017 10:51:00Grand GulfNRC Region 4GE-6

At approximately 0340 CDT on Wednesday, August 23, 2017, Grand Gulf Nuclear Station was notified by the Entergy System Dispatcher that the NRC had called them and told them that the NRC could not contact Grand Gulf on the Emergency Notification System (ENS) line nor commercial telephone. Control Room personnel immediately tested several offsite lines including the NRC ENS line and found the lines were non-functional. Offsite prompt Public Warning Sirens were available at all times. State and Local notification capability was available via UHF radio communication. GGN Emergency Response Organization notification capability was available at all times via satellite phone activation of group paging. GGN site Emergency Response Facility intercommunications were available at all times via site internal telephones. In-plant and offsite team communications were available at all times via UHF radio. This event is being reported (8-hour notification) as an event or condition that adversely impacted offsite communications in accordance with 10 CFR 50.72(b)(3)(xiii), specifically the loss of the NRC Emergency Notification System (ENS). The phone company has been contacted and actions are being taken to restore normal communications capability at this time. The NRC Resident Inspector has been notified.

  • * * UPDATE AT 1701 EDT ON 08/23/17 FROM LEROY PURDY TO JEFF HERRERA * * *

At 1701 EDT on 8/23/17 the phone systems at Grand Gulf have been restored. The licensee will be notifying the NRC Resident Inspector. Notified the R4DO (Farnholtz)

ENS 5291518 August 2017 23:41:00River BendNRC Region 4GE-6At 2055 CDT on August 18, 2017, an automatic actuation of the reactor protection system occurred while the plant was operating at 100 percent power. No plant parameters requiring the actuation of the emergency diesel generators or the emergency core cooling system were exceeded. The main feedwater system remained in service following the scram to maintain reactor water level, and the main condenser remained available as the normal heat sink. The scram occurred after a planned swap of the main feedwater master controller channels in preparation for scheduled surveillance testing. When the channel swap was actuated, the feedwater regulating valves moved to the fully open position. The scram signal originated in the high-flux detection function of the average power range monitors, apparently from the rapid increase in feedwater flow. The cause of the apparent feedwater controller malfunction is under investigation. The NRC Resident Inspector has been notified. No safety relief valves opened. Decay heat is being removed via steam to the main condenser using the bypass valves and steam drains. The licensee intends to go to Cold Shutdown to investigate the malfunction.
ENS 5290815 August 2017 20:21:00River BendNRC Region 4GE-6A blind sample provided from an independent laboratory to fleet testing facility was returned with inaccurate results. This is in violation of 10 CFR 26.719(c)(3) and requires a 24-hour report. The licensee has notified the NRC Resident Inspector.
ENS 5290315 August 2017 04:58:00PerryNRC Region 3GE-6On August 14th, 2017 at 2257 (EDT), while shutting down the Annulus Exhaust Gas Treatment System (AEGTS) Train B, secondary containment pressure momentarily lowered. This resulted in the Technical Specification (TS) for Secondary Containment to not be met for 15 seconds. The minimum Secondary Containment vacuum observed during that time was 0.52 inch of vacuum water gauge. Secondary Containment pressure was returned to within the TS operability limit of 0.66 inch of vacuum water gauge (TS SR 3.6.4.1.1) by the AEGTS Train A that remained in operation. There were no radiological releases associated with this event. Declaring Secondary Containment inoperable is reportable under (10 CFR) 50.72(b)(3)(v)(C) & (D) for an event or condition that could have prevented the fulfillment of a safety function of a system that is needed to:(C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector.
ENS 5289710 August 2017 17:14:00River BendNRC Region 4GE-6A non licensed supervisor confirmed positive for alcohol during a fitness for duty test. The employee's access has been terminated. The NRC Resident Inspector has been notified.
ENS 528918 August 2017 20:22:00PerryNRC Region 3GE-6On August 8, 2017, at 1554 hours (EDT), during restoration from testing of the High Pressure Core Spray (HPCS) Suppression Pool Level High Instrumentation, unexpected as-left indications were found that impacted both of the required channels of instrumentation. Subsequent venting of the instrumentation lines was completed and both channels of instrumentation are reading consistent with previously taken as-found data. The instrumentation was declared OPERABLE at 1635. The initial cause of the unexpected as-left indications appears to be the introduction of air into the instrumentation lines during the calibration activities. This is considered a loss of safety function based on both of the HPCS Suppression Pool Level High Instrumentation channels being declared INOPERABLE and the loss of the automatic HPCS suction swap to the Suppression Pool on a high level. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The (NRC Resident Inspector) has been notified.
ENS 5286217 July 2017 10:55:00ClintonNRC Region 3GE-6The following information is provided as a 60-day telephone notification to the NRC in accordance with 10 CFR 50.73(a)(1) reported under 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of the Division 3 emergency diesel generator (DG). The event occurred on May 18, 2017, at 1115 CDT. As allowed by 10 CFR 50.73(a)(1), the notification is being made via telephone. (a) The specific train(s) and system(s) that actuated were: During troubleshooting of blown fuses for the Reserve Auxiliary Transformer (RAT) main feed metering and relaying circuit, the Division 3 DG automatically started as a result of a loss of power signal, the RAT feed breaker for the offsite power source opened after a 15 second time delay as a result of a degraded voltage signal, and the DG output breaker subsequently closed. The loss of voltage and degraded voltage signals were generated when maintenance technicians opened the wrong test switch in the Division 3 4160-Volt Switchgear 1E22S004. (b) Whether each train actuation was complete or partial: Upon receiving the simulated loss of voltage and degraded voltage signals, the Division 3 DG started and the DG breaker closed as expected. No additional actuations occurred. (c) Whether or not the system started and functioned successfully: Upon receiving the simulated loss of voltage and degraded voltage signals, the Division 3 DG and the DG breaker were verified to have properly functioned in response to the invalid signals. The NRC Resident Inspector has been notified.
ENS 5284310 July 2017 05:42:00Grand GulfNRC Region 4GE-6At 2158 (CDT) a door to the Control Room Envelope was left unsecure. GGNS (Grand Gulf Nuclear Station) 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 and therefore could have prevented fulfillment of its safety function. The Control Room Envelope was inoperable for one minute. This event is being reported under 10CFR 50.72(b)(3)(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 licensee notified the NRC Resident Inspector.
ENS 5282523 June 2017 23:58:00River BendNRC Region 4GE-6While performing a scheduled generator voltage regulator test, River Bend Station experienced an automatic scram when the main generator tripped. It is unknown at this time why the main generator tripped. There were no equipment issues that materially impacted post scram operator response. The intention at this time is to go to cold shutdown while the cause of the trip is investigated. All rods inserted during the scram. Reactor water level is being maintained via normal feedwater with decay heat being removed via turbine bypass valves to the main condenser. The electrical grid is stable and supplying plant loads via the normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector.
ENS 5280615 June 2017 13:14:00ClintonNRC Region 3GE-6At 0958 hours (CDT), during planned surveillance testing of the Division 3 Shutdown Service Water (SX) subsystem, the Division 3 SX pump tripped for unknown reasons. The Division 3 SX subsystem was declared inoperable and in accordance with Technical Specification 3.7.2, Action A.1, the High Pressure Core Spray (HPCS) system was declared inoperable. Since the HPCS system is a single train safety system, this event is reportable under 10CFR50.72(b)(3)(v)(D). An investigation is underway to determine the cause of the SX pump trip. The NRC Resident has been notified.
ENS 5280011 June 2017 01:00:00ClintonNRC Region 3GE-6At 2256 CDT on 6/10/17, Clinton operators manually scrammed the reactor from 99 percent power due to a loss of feedwater heating. The scram was uncomplicated and the plant is stable and in mode 3. All rods inserted and decay heat is being removed by the condenser. All offsite power is available. The cause of the loss of feedwater heating is under investigation. The NRC Resident Inspector and the State of Illinois have been notified.
ENS 527822 June 2017 11:13:00ClintonNRC Region 3GE-6On 6/2/2017 at 0241 CDT, Clinton Power Station entered Mode 2 with secondary containment boundary doors propped open. Specifically, both doors for Reactor Water Cleanup (RT) 'B' pump room were propped open with welding cables routed through pump room doors to perform welding in the RT pump room. At 0300 CDT, a Senior Reactor Operator identified that the doors were propped open and Secondary Containment was declared inoperable. LCO 3.6.4.1 Required Action A.1 was entered to restore Secondary Containment to Operable in four hours. At 0324 CDT, the cabling for the welding machine was removed and the doors were closed. Investigation determined that authorization had been granted while in mode 4, when secondary containment was not required to be operable. The doors were propped open at the beginning of the shift, prior to the mode change to mode 2 (0241 CDT). This loss of secondary containment is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The NRC Resident Inspector has been notified.
ENS 5277730 May 2017 22:22:00ClintonNRC Region 3GE-6At 2038 (CDT), Clinton Power Station received an automatic RPS (Reactor Protection System)actuation. EOP-1 (Emergency Operating Procedure) was entered on RPV (Reactor Pressure Vessel) Level 3. The cause of the scram is unknown at this time. All systems responded appropriately following the scram and the plant is currently stable. Reactor level is being maintained by normal feedwater and decay heat is being removed to the main condenser via the steam dump bypass valves. The plant is in a normal shutdown electrical lineup. The plant main generator was synchronized to the electrical grid and the plant was conducting control rod scram time testing at the time of the reactor trip. The licensee notified the NRC Resident Inspector.
ENS 5275214 May 2017 16:09:00ClintonNRC Region 3GE-6At 0730 (CDT) on 5/14/2017, a visitor was working in the Protected Area (PA) on the turbine building roof and discovered a blue 12 ounce can of beer in their cooler. This was discovered when the visitor was removing items from their cooler into a larger community cooler. The visitor immediately notified their escort of the prohibited item. The escort then notified Security of the event. Security took possession of the item and the individual was escorted offsite. The individual stated when they packed their cooler at home they thought they had picked up a blue can of soda and did not notice it was a blue can of beer. This event is being reported per 10CFR26.719(b). The licensee notified the NRC Resident Inspector.
ENS 5275012 May 2017 08:20:00ClintonNRC Region 3GE-6At 0045 (CDT) on May 12, 2017, it was discovered that a Primary Containment local leak rate test performed on Main Steam Isolation Valves (MSIV) exceeded its acceptance criteria. During Modes 1, 2, and 3, Technical Specification Surveillance Requirement 3.6.1.3.9 requires MSIV leakage for a single MSIV line to be less than or equal to 100 standard cubic feet per hour (scfh) (47,195 sccm) and requires the combined leakage rate for all MSIV leakage paths to be less than or equal to 200 scfh (94,390 sccm) when tested at 9 psig. As-found for the 'D' MSIV line leakage is 53,921.61 standard cubic centimeter per minute (sccm) for the 'D' Inboard MSIV 1B21F022D and 59,698.8 sccm for the 'D' Outboard MSIV 1B21F028D. As-found combined MSIV min-path leakage is 102,463 sccm. This event is being reported as a condition of the nuclear power plant, including its principal safety barriers, being seriously degraded per 10 CFR 50.72(b)(3)(ii)(A) since the Primary Containment Isolation Valves leakage limits for MSIVs were exceeded. The NRC Resident Inspector has been notified.
ENS 527293 May 2017 15:04:00ClintonNRC Region 3GE-6At 0204 (CDT) on 5/3/2017, a facilities person was removing the trash bags from the garbage can in the restroom of the Administrative Building inside the Protected Area. While emptying the trash, they discovered a 100ml alcoholic beverage container in the trash. The container was empty, however, there was an odor of alcohol coming from the bottle. The item was turned over to the security department. The investigation identified the last time this trash bag had been changed out was on 5/2/2017 at 1530 (CDT). This event is being reported per 10CFR26.719(b). The licensee has notified the NRC Resident Inspector.
ENS 527273 May 2017 12:59:00PerryNRC Region 3GE-6On April 30, 2017, at 1818 (EDT), the main turbine steam bypass valve #1 partially opened. Power was incrementally lowered. 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) Turbine Stop Valve Closure and Control Valve Fast Closure trip functions were declared inoperable due to the opening of the bypass valve which affects the bypass setpoint for those RPS trip functions. With the loss of these RPS trip functions 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, this also resulted in a loss of safety function. Currently the bypass valve is closed and the RWL, EOC-RPT and RPS function are operable. Troubleshooting continues to determine the issue with the main turbine that caused the bypass valve to open. NRC Resident Inspector has been notified.
ENS 526634 April 2017 06:57:00Grand GulfNRC Region 4GE-6At 0010 (CDT), 04/04/2017, the reactor was manually scrammed from approximately 75 (percent) core thermal power due Condensate Storage tank level lowering to 24 feet. All control rods fully inserted and all systems actuated and operated as designed. No safety relief valves actuated. Reactor level and pressure are currently being controlled within normal bands. RCIC (reactor core isolation cooling) was manually initiated for level control. This event is reportable under 10CFR50.72(b)(2)(iv)(B) for the reactor trip and 50.72(b)(3)(iv)(A) for the manual start of the reactor core isolation cooling system. The cause of lowering level was a condensate pipe leak. Decay heat is being removed via steam dumps to the condenser. The electrical grid is stable and supplying plant loads. The licensee has notified the NRC Resident Inspector.
ENS 5263123 March 2017 07:24:00River BendNRC Region 4GE-6

River Bend Station personnel declared the High Pressure Core Spray (HPCS) system inoperable at 0256 on 3/23/2017. During performance of the HPCS Pump and Valve Operability Test, the operators observed an unusual system response after E22-MOVF023 (HPCS Test Return to the Suppression Pool) was stroked closed. A field check showed that the key that connects the E22-MOVF023 valve stem to the anti-rotation device had become dislodged. E22-MOVF023 is a Primary Containment Isolation Valve (PCIV) and is designed to close automatically on an ECCS (Emergency Core Cooling System) initiation signal to ensure that injection flow is directed to the reactor vessel. Technical Specification (TS) 3.6.1.3 requires that containment penetrations associated with an inoperable PCIV be isolated. E22-MOVF023 was declared inoperable at 0028. Operators were unable to close or demonstrate that E22-MOVF023 was fully closed as required by TS 3.6.1.3 and proceeded to isolate the associated containment penetration by closing other system valves. This action was completed at 0320. The net effect of the actions taken to isolate the containment penetration is that HPCS is inoperable as of 0256. This results in 14 day LCO. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM DAN JAMES TO KARL DIEDERICH ON 3/23/17 AT 10:01 EDT * * *

The Event Time was 0028 CDT rather than 0256 CDT. "The scheduled surveillance test of the high pressure core spray system was initiated at 2355 CDT on March 22, and the pump was secured at 0028 CDT on March 23. The inspection of the HPCS test return valve to the suppression pool occurred at 0050 CDT, and it was at that point that an apparent malfunction of the valve had occurred to the extent that it did not appear to be able to perform its safety function to close upon receipt of a design basis system initiation signal. Thus, the event time for this condition would be more accurately defined as 0028 CDT. Notified R4DO (James Drake) via e-mail.

ENS 5260210 March 2017 11:41:00River BendNRC Region 4GE-6At 0714 CST on March 10, 2017, with the unit in Mode 1 at approximately 17% power, a manual actuation of the reactor protection system (RPS) was initiated due to rising reactor pressure caused by the closure of the Main Turbine Control Valves (MTCV's). The cause of the closure of the MTCV's is under investigation. The unit is currently stable in Mode 3. All control rods inserted as expected; water level control is stable in the normal control band and reactor pressure is being maintained with steam line drains (aligned to the main condenser). The NRC Senior Resident Inspector has been notified.
ENS 526019 March 2017 10:57:00ClintonNRC Region 3GE-6On March 7, 2017, Division 2 Residual Heat Removal (RHR) system was inoperable due to a scheduled maintenance system outage window. At 2258 (CST), Operations identified a Division 1 Unit Substation Switchgear relay was cycling, which is part of the Division 1 AC Power system. The specific relay could not be identified at the time. Division 1 AC Power systems were protected. On March 8, 2017 at 1830 hours, Division 2 RHR was restored to operable status. On March 9, 2017 at 0319 hours, Operations declared Division 1 Emergency Diesel Generator (EDG) inoperable due to the (identification of the) Division 1 relay as related to properly tripping non-essential loads on a bus under-voltage condition. The relay would not have actuated to trip non-essential loads. The proper tripping of non-essential loads is a requirement for Division 1 EDG. The Updated Safety Analysis Report (USAR) Emergency Core Cooling Systems (ECCS) analysis specifies with the Division 1 DG failure, the remaining systems available are: Automatic Depressurization System (ADS), High Pressure Core Spray (HPCS), and 2 Low Pressure Core Injection (LPCI) systems. As a result of Division 2 RHR (being) inoperable at the same time Division 1 EDG was inoperable, an unanalyzed condition existed. While Division 2 RHR was inoperable, Division 1 EDG was inoperable. Technical Specification (TS) Limiting Condition of Operation (LCO) 3.8.1, AC Sources - Operating, was not met. Condition B, One Required DG Inoperable, Required Action B.2 declares required features, (normally) supported by the inoperable DG, inoperable when the redundant required features are inoperable, with a completion time of 4 hours. The action would have required declaring Division 1 ECCS inoperable, which includes Division 1 RHR and Low Pressure Core Spray (LPCS). With Division 1 EDG, Division 1 RHR, and Division 2 RHR inoperable, the station did not satisfy the USAR ECCS analysis and was in an unanalyzed condition. This condition is reportable under 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition, since the condition occurred within three years of the date of discovery. The NRC Resident Inspector has been notified.
ENS 5257625 February 2017 04:23:00ClintonNRC Region 3GE-6At approximately 2239 (CST) on 2/24/17, the Main Control Room received numerous annunciators that indicated a loss of the 138 kV off-site feed to the Emergency Reserve Auxiliary Transformer (ERAT). 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 running VF fans, secondary containment differential pressure rose to slightly greater than 0 inches water gauge which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge. The Control Room entered EOP-8, Secondary Containment Control. This event is being reported as a Condition that Could Have Prevented Fulfillment of a Safety Function under 10CFR50.72(b)(3)(v)(C). Secondary Containment differential pressure was restored within Technical Specification requirements at 2242 (CST) by starting the Standby Gas Treatment HVAC (VG) system. The NRC Resident (Inspector) has been notified.
ENS 5256820 February 2017 17:24:00River BendNRC Region 4GE-6

During the investigation associated with Event Notification 52566 that was reported on 2/18/17, it has been determined that an unanalyzed condition (new potential single failure concerns) exists. This condition exists only during periods of manually alternating divisions of Control Building Chilled Water systems; in that potential failures of Control Room Air Handling Units (HVC-ACU1A or B) or Control Building Air Handling Units (HVC-ACU2A or B) could fail in a manner that challenges the operability of the alternate division.

As reported in Event Notification 52566, the impact of this event was a loss of safety function cooling to both Division 1 and 2 AC/DC power distribution systems and Divisions 1 and 2 Control Room Fresh Air systems. Contingency actions are in development to address the impact of the potential failure mode. The plant remains in a planned refueling outage, Mode 5 with water level greater than 23' above the vessel flange. Shutdown cooling remains in service and is not affected by this issue. Investigation is ongoing. The NRC Resident Inspector has been briefed on this issue.

  • * * UPDATE FROM ROB MELTON TO DONALD NORWOOD AT 2129 EST ON 2/20/2017 * * *

The licensee updated information in the first paragraph of the original above with the following: During the investigation associated with Event Notification 52566 that was reported on 2/18/17, it has been determined that an unanalyzed condition (new potential single failure concerns) exists. During periods of alternating divisions of Control Building Chilled Water systems, the potential exists for failures of Control Room Air Handling Units (HVC-ACU1A or B) or Control Building Air Handling Units (HVC-ACU2A or B) that could challenge the operability of the alternate division. The licensee notified the NRC Resident Inspector of this update. Notified R4DO (Gepford)

  • * * UPDATE FROM STEVEN CARTER TO MARK ABRAMOVITZ AT 1513 EDT ON 2/22/17 * * *

After further investigation it has been determined that an unanalyzed condition (new single failure concerns) exists with the dampers associated with the Control Room Fresh Air system. The potential exists for damper failures for HVC-FN1A Control Room Booster Fan 1A motor and HVC-FN1B Control Room Booster Fan 1B motor that could challenge the operability of the alternate division. Contingency actions are in development to address the impact of the potential failure mode. The plant remains in a planned refueling outage, Mode 5, with water level greater than 23 feet above the vessel flange. Shutdown cooling remains in service and is not affected by this issue. Investigation is ongoing. The NRC Resident Inspector has been briefed on this issue. Notified R4DO (Pick).

ENS 5256619 February 2017 00:21:00River BendNRC Region 4GE-6At 1537 CST on February 18th, 2017, while the plant was in MODE 5 for a scheduled refueling outage, the main control room experienced a loss of Control Building chilled water and the associated ventilation systems while attempting to alternate divisions for testing. An equipment malfunction in a breaker supplying a Main Control Room air handling unit caused a loss of both divisions of Control Room and Control Building chilled water systems and associated ventilation systems until 1737 CST. During the period between 1537 and 1737, neither division of Control Building chilled water was able to perform the support function for cooling Division 1 and 2 AC and DC power distribution systems or the support function for the Division 1 and 2 Control Room Fresh Air systems. Shutdown Cooling remained in service throughout this event. There were no apparent effects on any plant equipment from the loss of chill water and ventilation. The Division 1 Control Building chill water and ventilation system was returned to service at 1737 on February 18, 2017. Actions were initiated to terminate the OPDRV (operations with potential to drain the reactor vessel) that was in progress at the time of the event by installing the reactor recirculation pump seal. As a conservative measure, actions were initiated to set containment and containment was set at 2145. Troubleshooting and analysis is ongoing to confirm and correct the problem which caused the loss of the Control Building chill water and ventilation system. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B). The NRC Senior Resident Inspector has been notified.
ENS 5255516 February 2017 13:03:00ClintonNRC Region 3GE-6On February 15, 2017 at 1515, it was discovered by corporate Fitness for Duty (FFD) personnel that an unescorted access reactivation feature in the security database (Illuminate) does not reset the flag to include an individual in the random FFD pool due to a database coding error. The Illuminate database was implemented fleet-wide 1/3/17. Review by corporate FFD personnel found one individual currently badged at Clinton Power Station was affected by the coding error. The individual was not in the FFD random pool from 1/3/17 until 2/15/17. Corporate security personnel found no other individuals to be affected by this issue. Affected individual was added to the FFD random pool. Corporate security personnel notified all Exelon sites of the issue. Sites were notified that the ability to use the re-activation feature in Illuminate would be removed from use by site personnel. Pending removal, a daily query would be run in the database to assure the re-activation feature had not been used by site personnel. The licensee informed the NRC Resident Inspector.
ENS 5251729 January 2017 16:30:00River BendNRC Region 4GE-6At 0209 CST, on January 29, 2017, while the plant was in MODE 4 for a refueling outage, the main control room crew removed the AC/DC inverter in the Division 1, 120 VAC electrical distribution system from service due to an equipment malfunction. Removing the inverter from service caused a loss of the associated 120 VAC instrument buss. This instrument buss loss caused a trip of the Division 1 Control Building Chill Water and Ventilation system. The Division 2 Control Building Chill Water and Ventilation System was locked out for surveillance testing at the time of the equipment failure. This condition rendered both divisions of Control Building Chill Water and Ventilation Systems unable to perform the support function for cooling Division 1 and 2 AC and DC power distribution systems. These systems are required to support the operability of two required divisions of shutdown cooling. Division 2 Shutdown Cooling System was in service and remained in service through out the event. The Division 2 Control Building Chill Water and Ventilation System was returned to service at 0220 CST on January 29, 2017. Division 1 Control Building Chill Water remains inoperable pending restoration with the installed backup Division 1 DC/AC inverter. Actions are ongoing to place this component in service and restore the associated 120 VAC instrument buss. The equipment malfunction was limited to the Division 1 inverter. The investigation of the inverter failure is ongoing. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B). The NRC Senior Resident Inspector has been notified.
ENS 5251627 January 2017 23:05:00Grand GulfNRC Region 4GE-6This notification is to report a loss of safety function in accordance with 10 CFR 50.72(b)(3)(v)(D). At approximately 1808 CST hours on Friday, January 27, 2017, the Grand Gulf Nuclear Station Unit 1 High Pressure Core Spray (HPCS) system was declared inoperable due to the trip of the HPCS Jockey Pump. At the time of discovery, Unit 1 was in Mode 2 and raising power in the source range to return to power operations. No other safety systems were inoperable at the time of this event. Investigation into the cause of the event is ongoing and the system will be returned to operational status prior to proceeding to Mode 1. The licensee has notified the NRC Resident Inspector.
ENS 5246830 December 2016 14:40:00PerryNRC Region 3GE-6On December 28, 2016 at 2119 EST, the Standby Liquid Control system (SLC) subsystem A was declared inoperable in accordance with the surveillance instruction for performance of a routine surveillance test. At 2229 EST, control room operators received an out-of-service alarm for the explosive-actuated injection valve for SLC subsystem B and declared subsystem B inoperable, thereby rendering both subsystems inoperable. With both subsystems inoperable, the SLC system was unable to fulfill its safety function. At 2335, the surveillance was completed and subsystem A was restored to operable status, which restored the ability for the system to fulfill its safety function. Troubleshooting determined that the cause for subsystem B inoperability was an intermittent electrical connection for the explosive-actuated injection valve. Repairs were conducted and the subsystem was restored to operable status on December 29, 2016 at 1708 EST. This issue was entered into the Corrective Action Program and during post reportability review, it was determined that this was a reportable event under 10 CFR 50.72(b)(3)(v)(A) for an event or condition that could have prevented the fulfillment of a safety function of a system that is needed to shut down the reactor and maintain it in a safe shutdown condition and under 10 CFR 50.72(b)(3)(v)(D) for a system that was unavailable for accident mitigation. The NRC Resident Inspector has been notified.