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	Discovered date	Reporting criterion	Title	Event description
ENS 53059	8 November 2017 00:10:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Unit 1 High Pressure Coolant Injection Inoperable	On November 7, 2017 at 1810 (CST), Unit 1 High Pressure Coolant Injection (HPCI), was manually isolated following failure of the remote turbine trip pushbutton to function. Unit 1 HPCI Operability Testing was in progress to the point of securing the HPCI turbine with the remote manual pushbutton. The pushbutton failed to trip the turbine resulting in operator action to lower the flow controller setpoint and isolating the HPCI steam line. HPCI remains isolated and is Inoperable pending resolution of the Turbine Trip circuitry. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D). The HPCI system is a single train system and the loss of HPCI could impact the plant ability to mitigate the consequences of an accident. The Reactor Core Isolation Cooling (RCIC) system was confirmed operable. The NRC Senior Resident Inspector has been notified.
ENS 52955	8 September 2017 16:30:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Hpci System Inoperable	On September 8, 2017 at 1130 hours CDT, Unit Two High Pressure Coolant Injection (HPCI) Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2-2354) failed to meet the Technical Specification Allowable Value during calibration testing. Technical Specification Table 3.3.5.1-1 Allowable Value (3.f) requires greater than or equal to 634 gpm (3.14 inches water column as required by procedure). HPCI was subsequently declared inoperable. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D). 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. The Reactor Core Isolation Cooling (RCIC) system was confirmed operable. Note: On September 8, 2017 at 1140 hours CDT, the HPCI Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2- 2354) was successfully recalibrated and HPCI was returned to Operable status. The NRC Senior Resident Inspector has been notified. * * RETRACTION AT 1216 EDT ON 10/19/17 FROM RYAN DECKER TO DONG PARK * * * The purpose of this notification today (10/19/17) is to retract the ENS Report made on September 8, 2017 at 1545 hours CDT (ENS Report #52955). Upon further investigation, it was determined that a surveillance procedure contained an overly restrictive statement that directed operators to immediately declare the High Pressure Coolant Injection (HPCI) system inoperable when the HPCI Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2-2354) fails. This statement was in conflict with existing Technical Specification (TS) 3.3.5.1, Condition E, that allows seven days to restore the HPCI FIS (instrument channel only) to an operable status prior to entry into TS 3.3.5.1, Condition H, which requires declaring HPCI inoperable immediately. Hence, during the period of FIS inoperability (i.e., 10 minutes), the HPCI system was not required to be declared inoperable in accordance with Technical Specifications. Therefore, based on this information, ENS Report # 52955 is being retracted. Note: On September 8, 2017 at 1140 hours CDT, the HPCI Minimum Flow Valve MO 2-2301-14 flow indicating switch (FIS 2-2354) was successfully recalibrated and HPCI was returned to Operable status. The NRC Resident Inspector has been notified. Notified R3DO (Daley).
ENS 52758	16 May 2017 00:18:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection System Declared Inoperable	On May 15, 2017 at 1918 hours (CDT), Unit Two High Pressure Coolant Injection (HPCI) Minimum Flow Valve MO 2-2301-14 failed to open as required by procedure and HPCI was declared inoperable. When the HPCI Turbine was tripped, the Minimum Flow Valve did not open when system flow reduced to the low flow setpoint. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D). 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. In accordance with Technical Specification 3.5.1 Condition G, the Reactor Core Isolation Cooling (RCIC) system was confirmed operable. This places the plant in a 14-day LCO action statement. The licensee has notified the NRC Resident Inspector.
ENS 52663	4 April 2017 05:10:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram Due to Condensate Leak	At 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 52648	29 March 2017 23:44:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Scram Initiated During Startup	At 1844 CDT on 3/29/2017, Unit 2 initiated a manual scram due to multiple rods inserting. At 1842 during Unit 2 start-up, Intermediate Range Monitor (IRM) 'G' drifted low. The operator adjusted the range down one position with no immediate reaction. At 1844, a spike on IRM 'G' caused a half scram on Reactor Protection System (RPS) 'A' trip system. The half scram was being reset after evaluating no trip condition was present. As the operator reset groups 2 and 3, a trip signal from IRM 'F' was received on the RPS 'B' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 2-AOI-100-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment Isolations Systems did not receive an actuation signal and performed as designed. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) Reactor Protection System(RPS) including reactor scram and reactor trip'. This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 52645	31 January 2017 19:25:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Report Due to Invalid Eccs Actuation Signal	The following report is made pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to an unintended initiation signal that occurred on January 31, 2017 with James A. FitzPatrick Nuclear Power Plant (JAF) in Mode 5 at zero (0) percent power. On January 31, 2017 at 1425 (EST) the control room received multiple annunciations associated with the following Systems / Trains: Primary Containment Isolation System (PCIS) / Trains A and B Residual Heat Removal System (RHR) / Trains A and B Core Spray (CS) / Trains A and B Reactor Core Isolation Cooling (RCIC) All four (4) Emergency Diesel Generators (EDG) auto-started with their associated Emergency Service Water pumps operating. RHR and CS both received initiation signals but were defeated per procedure. The HPCI (High Pressure Coolant Injection) auxiliary oil pump was taken to Pull-to-Lock per procedure, and the RCIC steam isolation valve cycled until the breaker was opened to close the valve. An evaluation concluded that the (Emergency Core Cooling System - ECCS) initiation signals were caused by the opening of a portable job box that was stored near sensitive equipment. Upon opening the job box, the lid bumped a reference leg resulting in the initiation signals. All initiation signals were reset and systems restored to normal shutdown lineups. The licensee notified the NRC Resident Inspector. * * UPDATE ON 3/30/17 AT 0840 EDT FROM DUSTIN SCURLOCK TO DONG PARK * * * To the original report, the licensee added, "This condition recurred at 1624 (EDT on 1/31/17). The licensee notified the NRC Resident Inspector. Notified R1DO (Cook).
ENS 52643	27 March 2017 22:25:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Hpci Declared Inoperable Due to Inadvertent Isolation	On March 27, 2017, at 1825 hours EDT, with the reactor at 100 percent core thermal power and steady state conditions, technicians inadvertently caused a High Pressure Coolant Injection (HPCI) System isolation, by testing the incorrect temperature switches in the TIP (Traversing In-core Probe) room. Pilgrim Nuclear Power Station (PNPS) was performing testing on the temperature switches for Reactor Core Isolation Cooling (RCIC), but the HPCI temperature switches were inadvertently actuated causing HPCI to isolate. The Limiting Condition for Operation (LCO) Action Statement 3.5.c.2 has been entered and the planned testing has been secured pending further investigation. PNPS is providing an 8-hour non-emergency notification that the HPCI System was declared inoperable in accordance with 10 CFR 50.72(b)(3)(v)(D), an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. HPCI was returned to Operable within 40 minutes. The licensee notified the NRC Resident Inspector and the Commonwealth of Massachusetts.
ENS 52543	8 February 2017 16:51:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection System Inoperable Due to Degraded Dc to Ac Inverter	During a control room panel walk down by an on-shift Reactor Operator at approximately 1151 (EST) on 2/8/2017, Unit 1 High Pressure Coolant Injection (HPCI) suction and discharge pressure indicators were noted to be downscale. I & C investigated and found the output of 1E41K603, DC to AC inverter, degraded. This inverter also powers the HPCI flow controller. Without the flow controller HPCI would not auto-start to mitigate the consequences of an accident, thus HPCI was declared inoperable. All other emergency core cooling systems and the Reactor Core Isolation Cooling (RCIC) system remain operable. HPCI is a single train system with no redundant equipment in the same system, thus this failure is reportable as an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident, 10CFR50.72(b)(3)(v)(D). Inverter 1E41K603 was replaced and functionally tested satisfactorily at 1630 on 2/8/2017, restoring HPCI to operable status. The NRC Resident Inspector was notified.
ENS 52442	18 December 2016 19:24:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Scram Due to Load Reject from Substation	On December 18, 2016 at time 1124 PST the plant experienced a full reactor scram. Preliminary investigations indicate that the scram was caused by a load reject from the Bonneville Power Administration (BPA) Ashe substation. Further investigations continue. The following conditions have occurred: Turbine Governor valve closure Reactor high pressure trip +13 inches reactor water level activations E-TR-B (backup transformer) supplying E-SM-7/SM-8 (vital power electrical busses) Complete loss of Reactor Closed Cooling (RCC) E-TR-S (Startup transformer) supplying SM-1/2/3 (non-vital power electrical busses) E-DG-1/2/3 (emergency diesel generators) auto start Low Pressure Core Spray (LPCS) and Residual Heat Removal (RHR) A/B/C initiation signals Main Steam Isolation Valves (MSIV) are closed Reactor Core Isolation Cooling (RCIC) RCIC and High Pressure Core Spray (HPCS) were manually activated and utilized to inject and maintain reactor water level. Pressure control is with Safety Relief Valves (SRV) in, manual. Level control is with RCIC and Control Rod Drive (CRD). RCIC has experienced an over speed trip that was reset so that level control could be maintained by RCIC. This event is being reported under the following: 10 CFR 50.72(b)(2)(iv)(A) which requires a 4 hour notification for Emergency Core Cooling System (ECCS) discharge into the reactor coolant system. 10 CFR 50.72(b)(2)(iv)(B) which requires a 4 hour notification for any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical. 10 CFR 50.72(b)(3)(iv)(A) which requires an 8 hours notification for actuation of ECCS systems. All control rods fully inserted. The NRC Resident Inspector has been informed. The licensee indicated that no increase in radiation levels were detected.
ENS 52419	9 December 2016 03:37:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Hpci Surveillance Failure	On 12/08/16 at approximately 2237 (EST), the Unit 2 HPCI (High Pressure Coolant Injection) system failed to meet surveillance testing requirements for achieving rated flow at greater than or equal to a minimum test pressure established per the surveillance. Operations declared the HPCI system inoperable and entered Technical Specification 3.5.1 Condition C for HPCI being inoperable. Other standby systems (Reactor Core Isolation Cooling and low pressure emergency core cooling systems) are operable. HPCI is a single train system. Therefore, per NUREG-1022, this condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of the safety function of a system required to mitigate the consequences of a design event. This condition has been entered into the Corrective Action program (IR 3951006). Investigation of the exact failure condition is in progress so that repairs can be made. At the surveillance flow of 5,000 gpm, the system was approximately 80 psi below the required pressure of 1,278 psi. Technical Specification 3.5.1, Condition C, is a 14-day Limiting Condition of Operation. The NRC Resident Inspector will be notified. * * RETRACTION AT 1440 EST ON 01/19/17 FROM ELMER KAUFFMAN TO S. SANDIN * * * The licensee provided the following information as the basis for retracting this report: This is a retraction of an event notification made on 12/09/16 at 0529 EST (EN #52419). This event was initially reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that, at the time of discovery, was believed to have prevented the fulfillment of the High Pressure Coolant Injection (HPCI) system safety function. On 12/08/16 at 2237 EST, the Unit 2 HPCI system was declared inoperable due to failing to meet surveillance testing requirements for achieving rated flow at greater than or equal to a minimum test pressure established per the surveillance. Prompt troubleshooting was performed and it was determined that an adjustment to the HPCI turbine governor control system was required. This adjustment was performed and HPCI was returned to an operable status on 12/09/16. Subsequent to this occurrence, Engineering has completed an evaluation that concluded that HPCI was capable of fulfilling its safety function and that the associated Technical Specification (TS) Surveillance Requirement (SR) 3.5.1.8 was met. The evaluation concluded that HPCI was degraded, but met the threshold for TS operability. The NRC Senior Resident has been informed of this retraction." Notified R1DO (Kennedy).
ENS 52334	31 October 2016 07:39:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection (Hpci) Inoperable	On October 31, 2016, at 0239 hours (CDT), a defect (minor audible through-wall leak) was identified on the steam line drain valve 1-2301-55, HPCI Steam Line Drain Line Steam Trap Outlet Valve. The defect was identified by Operations personnel traversing through the HPCI room as part of normal rounds. HPCI was declared inoperable under Tech Specs 3.5.1, Condition G. The Reactor Core Isolation Cooling (RCIC) system was verified operable. HPCI remains available (but not operable). The leak has been isolated. The 1-2301-55 is a manual valve downstream of the HPCI steam line drain trap. In a standby line-up, this line drains condensation from the HPCI steam supply line to the main condenser. During operation in an accident scenario, this line drains condensation from the HPCI steam supply line to the Torus via a drain pot. The location of the defect is in class 2 safety related piping. HPCI is a single train safety system and this notification is being made in accordance with 10CFR50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified. Technical Specification 3.5.1, condition G requires that HPCI be Operable within 14 days. * * RETRACTION ON 12/05/2016 AT 1505 EST FROM MARK BRIDGES TO STEVEN VITTO * * * The purpose of this notification is to retract the ENS Report made on October 31, 2016, at 0239 hours CDT (ENS Report #52334). Upon further investigation, a pinhole through-wall leak was discovered in the body of the 1-2301-55 valve (HPCI Steam Line Drain Line Steam Trap Outlet Valve). The defect was characterized as a 1/32-inch rounded hole due to a manufacturing defect in the casting located on the downstream side of the valve near the piping connection. A subsequent evaluation performed by Quad Cities Station considering the defect size, location, and characterization, confirmed the Unit 1 High Pressure Coolant Injection (HPCI) system would have performed its safety function when required. Based on this subsequent evaluation, ENS Report 52334 is being retracted. Note: On November 1, 2016, at 1624 hours CDT, the 1-2301-55 valve (HPCI Steam Line Drain Line Steam Trap Outlet Valve) was successfully repaired and HPCI was returned to Operable status. The NRC Resident Inspector has been notified. Notified R3DO (Stone).
ENS 52264	26 September 2016 22:45:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection System Declared Inoperable	On 9/26/16 at approximately 1845 EDT, investigation of an identified water leak on one of the two Unit 3 HPCI turbine exhaust drains to the drain pot determined that there was through wall leakage of approximately 2 drops per minute. Operations promptly declared the HPCI system inoperable and entered Technical Specification 3.5.1 Condition C for HPCI being inoperable. Other standby systems (Reactor Core Isolation Cooling and low pressure emergency core cooling systems) are operable. HPCI is a single train system. Therefore, per NUREG-1022, this condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of the safety function of a system required to mitigate the consequences of a design event. This condition has been entered into the corrective action program (IR 2720241). Investigation of the exact flaw location is in progress so that repairs can be made. The NRC resident has been informed of this notification.
ENS 52159	7 August 2016 05:01:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection System Declared Inoperable	HPCI (high pressure coolant injection) governor valve did not respond as expected. During performance of a planned HPCI valve functional test the HPCI governor valve (FD-FV-4879) did not reposition as expected. The HCPI system has been declared inoperable based on the response per Technical Specification 3.5.1 action C.1. All other emergency core cooling systems and the reactor core isolation cooling (RCIC) system remain operable. The unit remains at 100% power. The station has initiated an event response team to identify and correct the cause of the failure. No personnel injuries resulted from the event. The licensee notified the NRC Resident Inspector and the Lower Alloways Creek Dispatch.
ENS 52064	5 July 2016 20:40:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection (Hpci) System Declared Inoperable	On July 5, 2016, at 1640 Eastern Daylight Savings Time (EDT) the Unit 2 HPCI system was declared inoperable due to apparent failure of the HPCI Auxiliary Oil Pump after the 'HPCI Aux Oil Pump Motor Overload' control room annunciator was received. Failure of the HPCI Auxiliary Oil Pump prevents the HPCI system from performing its design safety function. As such, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. This event did not result in any adverse impact to the health and safety of the public. The NRC Senior Resident Inspector has been notified. The safety significance of this condition is minimal. All other Emergency Core Cooling Systems and the Reactor Core Isolation Cooling (RCIC) system remain operable. Troubleshooting activities are in progress. The HPCI system will remain inoperable until the cause of the failure has been corrected.
ENS 51715	7 February 2016 18:26:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Scram and Alert Declaration Due to Electrical Fault Resulting in Fire/Explosion	At 1346 EST the licensee reported that at 1326, Brunswick Unit 1 declared an Alert under EAL HA 2.1 due to an explosion/fire in the Balance of Plant 4 kV switchgear bus area. Prior to the Alert declaration, the operators initiated a manual SCRAM due to an unexpected power decrease from 88% to 40%. The licensee has visually verified that there is no ongoing fire and is investigating the initial cause of the event. Offsite power is available to the site, but EDGs 1 and 2 are running and supplying Unit 1 loads. The MSIVs shut and HPCI/RCIC are being used to maintain vessel level. At 1412 EST, NRC decided to remain in Normal Mode. At 1704 EST the licensee reported the following: At 1313 hours Eastern Standard Time (EST) a manual reactor scram was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. At this time, a Startup Auxiliary Transformer (SAT) experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. Emergency Diesel Generators (EDGs) 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The power interruption resulted in closure of the Main Steam Isolation Valves, per design. The manual scram also resulted in closure of Group 2, 6, and 6 Containment Isolation Valves. The Reactor Core Isolation Cooling (RCIC) system was manually started and is being used to control reactor water level. The High Pressure Coolant Injection (HPCI) system was manually started and is being used for pressure control. The Plant response to the event was per design. Unit 2 is not directly affected by the event, however, due to the shared electrical distribution system is in a Technical Specification Action statement due to the Inoperable Unit 1 SAT. The public health and safety is not impacted by this event. At 1751 EST, the licensee reported that the emergency declaration had been downgraded to an Unusual Event at 1730 because the plant no longer meets the criteria for an Alert, but does meet the criteria for an Unusual Event due to a "loss of all offsite power to Emergency 4 kV buses E1 (E3) and E2 (E4) for greater than or equal to 15 minutes." The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email). * * UPDATE FROM MARTY IRWIN TO DANIEL MILLS AT 1825 ON 2/07/16 * * * At 1814 EST the emergency declaration was terminated because offsite power was restored. The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified R2DO (Musser), NRR EO (Morris), IRD MOC (Stapleton), R2RA (Haney), NRR ET (Lubinski), NRR ET (Dean), DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
ENS 51630	31 December 2015 11:30:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection Inoperable	On 12/31/15, at approximately 0630 hours (EST), during shift turnover panel walk-downs, a licensed Unit 3 reactor operator identified that the High Pressure Coolant Injection (HPCI) flow controller output indicated a downscale condition. The controller was in automatic with the set point at 5000 gpm, which would typically indicate a controller output value of 100%. HPCI was not in operation and is a standby system. Operations promptly declared the HPCI system inoperable and entered Technical Specification Condition C for HPCI being inoperable. Other standby systems (Reactor Core Isolation Cooling and low pressure emergency core cooling systems) are OPERABLE. HPCI is a single train system. Therefore, per NUREG-1022, this condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of the safety function of a system required to mitigate the consequences of a design event. This condition has been entered into the Corrective Action program (IR 2606215). Maintenance troubleshooting of the flow controller loop has identified a failed component and repair activities are in progress. The NRC Resident Inspector has been informed of this notification.
ENS 51391	14 September 2015 03:05:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Manual Scram Due to Loss of Turbine Building Closed Cooling Water	At 2305 EDT on September 13, 2015, a manual scram was initiated in response to a loss of all Turbine Building Closed Cooling Water (TBCCW). All control rods fully inserted. The lowest Reactor Water Level (RWL) reached was 137 inches. All isolations and actuations for RWL 3 occurred as expected. Decay heat was initially being removed through the Main Turbine Bypass System to the Main Condenser, however, as a result of the loss of TBCCW, the Main Feed Pumps lost cooling and had to be secured. At 2310, Standby Feedwater was initiated and Main Feedwater was secured. The loss of TBCCW also caused all Station Air Compressors (SACs) to trip on loss of cooling. The loss of SACs caused the Instrument Air header pressure to degrade to the point at which the Secondary Containment isolation dampers drifted closed. This resulted in the Reactor Building vacuum exceeding the Technical Specification limit. At 2325, operators started the Standby Gas Treatment system and manually initiated a Secondary Containment isolation signal. Secondary Containment vacuum was promptly restored to within Technical Specification limits. Additionally, Operators were monitoring for expected MSIV drift due to the degraded Instrument Air header pressure. When outboard MSIVs were observed to be drifting, Operators closed the outboard and inboard MSIVs at 2345. At 2352, Safety Relief Valves (SRVs) reached the Low-Low Setpoint and began cycling to control reactor pressure. RWL is currently being maintained in the normal level band with the Standby Feedwater and Control Rod Drive systems. Reactor Pressure is being controlled with Safety Relief Valves. Operators are currently in the Emergency Operating Procedure for Reactor Pressure Vessel control. Investigation into the loss of TBCCW continues. No safety-related equipment was out of service at the time of the event. All offsite power sources were adequate and available throughout the duration of the event. The NRC resident inspector has been notified. * * UPDATE AT 0555 EDT AT 09/14/15 FROM CHRIS ROBINSON TO JEFF HERRERA * * * At 0409 EDT the Reactor Core Isolation Cooling (RCIC) system was placed in service due to identification of an unisolable leak in the Standby Feedwater System. Reactor water level and pressure is now being controlled though the RCIC system and Safety Relief Valves. This event update is reportable as a valid manual initiation of a specified safety system under 10CFR50.72(b)(3)(iv)(A). The NRC resident inspector has been notified. The leak rate was reported as approximately 5-10 gallons per minute from a weld on the standby feedwater pump header drain valve F326. The licensee reported the leak stopped once RCIC was placed into service. The licensee is still investigating the issue. Notified the R3DO (Pelke), IRD Manager (Grant), NRR EO (Morris). * * UPDATE PROVIDED BY CHRIS ROBINSON TO JEFF ROTTON AT 2135 EDT ON 09/14/2015 * * * At 1847 EDT on September 14, 2015, a valid automatic Reactor Protection System (RPS) actuation occurred due to Reactor Water Level 3 while shutdown in MODE 3. Operators were manually controlling Reactor Pressure Vessel (RPV) level and pressure with Reactor Core Isolation Cooling (RCIC) and Safety Relief Valves (SRV). While operators were cycling SRVs, the RPV level went below the Level 3 setpoint. Operators promptly restored RPV level by manual operation of RCIC. The Level 3 actuation and associated isolations were verified to operate properly. The scram signal has been reset. Fermi 2 remains in MODE 3 controlling RPV Level and Pressure through manual operation of RCIC and SRVs. This is the second occurrence of a valid specified safety system actuation reportable under 10CFR50.72(b)(3)(iv)(A) for this ongoing event. The NRC Resident Inspector has been notified. Notified R3DO (Riemer), IRD Manager (Grant), and NRR EO (Morris) * * UPDATE FROM BRETT JEBBIA TO JOHN SHOEMAKER AT 1446 EST ON 2/27/16 * * * This update provides clarification of the applicable reporting criteria for this Event associated with primary containment isolation actuations. Upon the manual reactor scram at 2305 EDT on September 13, 2015, Reactor Protection System (RPS) Level 3 actuated and Primary Containment Isolation System (PCIS) Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for these actuations is 10 CFR 50.72(b)(3)(iv)(A). The applicable reporting criterion for the manual closure of the inboard and outboard main steam isolation valves at 2345 EDT on September 13, 2015, is also 10 CFR 50.72(b)(3)(iv)(A). In addition, the manual closures of all MSIV lead to a loss of condenser vacuum which resulted in the actuation of PCIS Group 1 at 0001 EDT on September 14, 2015, as expected. The applicable reporting criterion for this actuation is also 10 CFR 50.72(b)(3)(iv)(A). Upon reaching Level 3 at 1847 EDT on September 14, 2015, PCIS Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for this actuation is 10 CFR 50.72(b)(3)(iv)(A). The licensee informed the NRC Resident Inspector. Notified the R3DO (Stone).
ENS 50956	5 April 2015 19:37:00	10 CFR 50.72(a)(1)(i), Emergency Class Declaration	Alert Declared Due to Fire in Motor Control Center	An Alert was declared due to a fire in a Unit 2 Division 2 Safeguard (250 volt) DC Motor Control Center. This has made the High Pressure Core Injection system inoperable and unavailable. The fire is out. The emergency response organization has been activated and investigation / repair planning will commence. Unit 2 is stable with no other system affects. The fire was extinguished by on-site personnel. No off-site responders were required. The Reactor Core Isolation Cooling system remains operable. A Fire Watch has been stationed to monitor for fire reflash. There were no injuries resulting from this event. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector and State and local government agencies. Notified the following organizations: DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, FDA EOC, FEMA NWC, and Nuclear SSA. * * UPDATE FROM DAN BOYLAN TO DONALD NORWOOD AT 1812 EDT ON 04/05/15 * * * The Alert was terminated at 1742 EDT. The licensee notified the NRC Resident Inspector and State and local government agencies. Notified R1DO (Kennedy), IRD (Gott), and NRR EO (Morris). Notified the following organizations: DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, FDA EOC, FEMA NWC, and Nuclear SSA.
ENS 50816	12 February 2015 18:36:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Hpci Declared Inoperable During Weekly Inspection	EVENT DESCRIPTION: On February 12, 2015, at 1336 Eastern Standard Time (EST) the Unit 1 High Pressure Coolant Injection (HPCI) system was declared inoperable due to a failure of the HPCI Auxiliary Oil Pump. During performance of a routine HPCI weekly inspection, the auxiliary oil pump was started and subsequently experienced a loss of discharge oil pressure. The HPCI Auxiliary Oil Pump provides hydraulic pressure required to open the HPCI Turbine Stop Valve and the HPCI Turbine Control Valve during initial HPCI startup. Failure of the HPCI Auxiliary Oil Pump prevents the HPCI system from performing its design safety function. As such, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. This event did not result in any adverse impact to the health and safety of the public. INITIAL SAFETY SIGNIFICANCE EVALUATION: The safety significance of this condition is minimal. All other Emergency Core Cooling Systems and the Reactor Core Isolation Cooling (RCIC) system remain operable (per the requirements of 14-day LCO (Limiting Condition of Operation) 3.5.1). CORRECTIVE ACTIONS: Troubleshooting activities are in progress. The HPCI system will remain inoperable until the cause of the failure has been corrected. The NRC Resident Inspector has been notified.
ENS 50771	27 January 2015 14:48:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Loss of High Pressure Coolant Injection	On Tuesday, January 27, 2015, at 0948 EST, with the Reactor Mode Select Switch (RMSS) in the Shutdown position and Pilgrim Nuclear Power Station (PNPS) at 0% core thermal power, the High Pressure Coolant Injection (HPCI) system was isolated by the main control room operating crew and declared INOPERABLE. HPCI had been in service for reactor pressure control following the automatic reactor scram experienced during winter storm 'Juno' reported in EN# 50769. It appears there was a malfunction of the HPCI turbine gland seal condenser blower or associated condensate pump. Reactor pressure control was transitioned to the safety relief valves and the reactor cooldown was continued. The plant is stable. The Emergency Diesel Generators are powering the safety related 4KV buses and reactor water level is being maintained by the Reactor Core Isolation Cooling (RCIC) system. HPCI is required to be OPERABLE in accordance with Technical Specification 3.5.C.1. Since HPCI is a single train system, the INOPERABILITY is reportable in accordance with 10CFR50.72(b)(3)(v)(D). The cause of the HPCI malfunction is not known at this time and troubleshooting continues. This event had no impact on the health and/or safety of the public. The USNRC Senior Resident Inspector has been notified. Shutdown cooling is in service.
ENS 50769	27 January 2015 09:02:00	10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram on Turbine Trip Due to Loss of Offsite Power	On Tuesday January 27, 2015 at 0402 hours, with the Pilgrim Nuclear Power Station (PNPS) Reactor Mode Select Switch (RMSS) in Run and reactor power approximately 52% an automatic reactor scram signal was received due to the automatic trip of the main turbine that was initiated by the opening of the main generator breaker, ACB-104. The event occurred during winter storm 'Juno.' Prior to the event off-site transmission Line 355 was de-energized due (to) weather conditions and its associated PNPS switchyard breakers (ACB-105, a main generator breaker and ACB-102), were open. Per station procedures, reactor power was being lowered, a reactor protection system bus had been placed onto a back-up power supply, the Emergency Diesel Generators had been started and were powering the associated safety related 4 KV buses. The second off-site transmission Line 342 de-energized and the associated PNPS switchyard breakers (ACB-104 main generator breaker and ACB-103) opened. The Shutdown Transformer off-site power supply has remained available throughout this event. All control rods were verified to be fully inserted. Per plant design, Primary Containment Isolation System (PCIS) Group lI sampling systems, Group VI Reactor Water Clean-up (RWCU) system and Reactor Building Isolation System (RBIS) isolations occurred. Currently, the EDG's are powering the safety related 4KV buses, reactor water level is being maintained by the Reactor Core Isolation Cooling (RCIC) system and reactor pressure is being maintained by High Pressure Coolant Injection (HPCI) system. The station is conducting a plant cool down at this time. All systems responded as designed with the exception of a non-safety-related diesel air compressor, K-117 that failed to start. The licensee will notify the State and local governments and plans on issuing a press release. The NRC Resident Inspector has been informed.
ENS 50675	12 December 2014 21:00:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition with Reactor Core Isolation Cooling	Engineering identified fuse and breaker coordination issues with Reactor Core Isolation Cooling (RCIC) valves operated at the Remote Shutdown Panel (RSDP). The coordination issues are such that, given a fire in the main control room, it is possible that RCIC valve power supply breakers could trip prior to tripping control power fuses. Operation of RCIC from the RSDP could be impaired in this scenario without compensatory actions to reset breakers. RCIC is the single credited source of makeup to the reactor pressure vessel during this scenario. The current licensing basis (Fire Protection Report) does not identify the compensatory actions required to reset breakers prior to RCIC operation at the RSDP. This condition is applicable to Unit 1 and Unit 2. This report is being made pursuant to 10CFR50.72(b)(3)(ii)(B), 'Event or Condition that results in an unanalyzed condition that significantly degrades plant safety'. Actions are being taken to amend the appropriate operating procedures to take the required steps to ensure proper operation of RCIC in the postulated scenario. The licensee has notified the NRC Resident Inspector.
ENS 50601	7 November 2014 13:47:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Reactor Scram Due to Loss of Feedwater	The Perry Nuclear Power Plant experienced an automatic reactor scram due to a loss of feedwater, which resulted in receiving valid reactor vessel water Level 3 and Level 2 initiation signals. The High Pressure Core Spray system and the Reactor Core Isolation Cooling system started and injected. Reactor water level and pressure have been stabilized in the required bands. The motor feed pump automatically started and is being used to control reactor vessel water level. The High Pressure Core Spray and Reactor Core Isolation Cooling systems have been returned to the standby mode. As a result of receiving a reactor vessel water Level 2 signal a Balance of Plant containment isolation signal was received. All systems isolated as required and the plant is restoring isolated systems in accordance with plant procedures. During the scram, all rods fully inserted into the core. Decay heat is being removed via turbine bypass valves to the main condenser. The electrical grid is stable and is supplying plant loads. An emergency diesel generator (Division 3 High Pressure Core Spray) started, as designed, as a result of the reactor vessel water Level 2 signal. No safety relief valves lifted as a result of the transient. The plant is stable with cooldown and depressurization to Mode 4 in progress. The cause of the loss of feedwater is under investigation. The NRC Resident Inspector has been notified. The State of Ohio and local officials will be notified.
ENS 50551	20 October 2014 06:18:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Reactor Scram on Loss of Feedwater	The Perry Power Plant experienced a reactor scram during a shift of non-essential vital power supply to the alternate source. Feedwater was lost resulting in receiving a valid level 3 and level 2 signal. High Pressure Core Spray and Reactor Core Isolation Cooling started and injected. Reactor level and pressure have been stabilized to required bands. The motor feed pump has been started and is controlling level. High Pressure Core Spray and Reactor Core Isolation Cooling have been returned to standby. During the scram, all rods fully inserted into the core. Decay heat is being removed via the steam dumps to the condenser. The electrical grid is stable and supplying plant loads. An emergency diesel generator started, as designed, as a result of the level 2 signal but did not load. No safety valves lifted as a result of the transient. The cause of the loss of feedwater is under investigation. The licensee will be notifying the State of Ohio and Perry Township and has notified the NRC Resident Inspector. * * UPDATE FROM DOUG SHORTER TO HOWIE CROUCH AT 0933 EDT ON 10/20/14 * * * The plant is currently in Mode 3, stable with cooldown and depressurization to Mode 4 in progress. Level control is being provided by the motor feedwater pump. Troubleshooting of the cause of the scram and loss of feed water is on-going. The initial notification identified 10CFR50.72(b)(3)(iv)(A), 'Specified System Actuation', as a reporting criteria. The specific system that actuated was not provided. As a result of receiving a reactor vessel water level 2 signal a containment/BOP isolation signal was received. All systems isolated as required and the plant is restoring isolated systems in accordance with procedure. The licensee will be notifying the State of Ohio and Perry Township and has notified the NRC Resident Inspector. Notified R3DO (Pelke).
ENS 50546	17 October 2014 08:03:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram on High Average Power Range Monitor Flux	At 0303 (CDT), River Bend Nuclear Station sustained a reactor scram due to high Average Power Range Monitor (APRM) flux, suspected due to a malfunction of the Electrohydraulic Control System. Reactor recirculation pump 'B' tripped, reactor recirculation pump 'A' responded appropriately. All other systems responded appropriately except for loss of feed water due to low suction pressure trip from isolating the condensate demineralizers. Reactor water level did not get out of level band. Condensate demineralizers and feedwater were restored to service. Level 3 (isolation) was initiated due to scram. (One) system, Suppression Pool Cooling isolated accordingly due to level 3 signal. Currently the plant is in mode 3, hot shutdown. Plant will remain in mode 3 until investigation of scram is complete. During the scram, all rods inserted into the core. No relief valves lifted as a result of the transient. All safety equipment is available although reactor core isolation cooling is functional but inoperable due to an earlier issue discovered during a surveillance test. The reactor is at normal pressure and temperature for Mode 3. The cause of the high APRM flux and the identified anomalies are under investigation. The licensee has notified the NRC Resident Inspector. * * UPDATE FROM DANIEL PIPKIN TO DANIEL MILLS AT 1043 EDT ON 10/17/2014 * * * The licensee is updating the notification to include an 8 hour notification for a specified system actuation due to the Level 3 isolation signal. Licensee is proceeding to cold shutdown to troubleshoot the EHC system. The licensee will notify the NRC Resident Inspector. Notified R4DO (Haire).
ENS 50404	26 August 2014 22:30:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Turbine Generator Neutral Overvoltage Causes a Reactor Scram	At 1730 CDT on August 26, 2014, Browns Ferry Unit 1 experienced a turbine trip resulting in an automatic reactor scram. The cause of the turbine trip was a control valve fast closure signal that was generated by a turbine trip on generator neutral over voltage signal. The Main Steam Isolation Valves (MSIVs) remained open with the main turbine bypass valves controlling reactor pressure. The Reactor Feedwater Pumps are in service to control reactor water level. Primary Containment Isolation Systems (PCIS) Groups 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required with the exception of Standby Gas Treatment (SBGT) train A, which is under a clearance for planned maintenance. Neither High Pressure Coolant Injection (HPCI) nor Reactor Core Isolation Cooling (RCIC) initiation signals were received. Initially, three Main Steam Relief Valves (MSRVs) opened to control the pressure surge and subsequently reclosed. This event requires a 4 hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including reactor scram or reactor trip, and (2) General containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).' The NRC Resident Inspector has been notified. Service Request 926468 was initiated in the Corrective Action Program. The plant is in its normal shutdown electrical lineup. The licensee is investigating the cause of the generator neutral overvoltage signal. There was no impact on units 2 and 3.
ENS 50346	5 August 2014 22:34:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram	This notification is being provided pursuant with SAF 1.6 10CFR50.72(b)(2)(iv)(B) and SAF 1.7 10CFR50.72(b)(3)(iv)(A). At 1734 CDT on August 5, 2014, LaSalle Unit 2 automatically scrammed due to an RPS actuation. The MSIVs isolated on a Group 1 signal, the cause is under investigation. The reactor water cleanup system isolated during the transient. The plant is stable with Reactor Pressure Control being maintained by the Reactor Core Isolation Cooling System and SRVs and level being controlled by the Low Pressure Core Spray System. The plant is planned to remain in hot shutdown pending investigation of the trip." The Unit 2 electric plant is in a normal shutdown lineup. All control rods inserted fully on the scram. Unit 1 was not affected by the Unit 2 transient. The licensee notified the NRC Resident Inspector. * * UPDATE PROVIDED BY MICHAEL FITZPATRICK TO JEFF ROTTON AT 1650 EDT ON 8/6/2014 * * * The initial notification to the NRC stated that the reactor water cleanup system had isolated during the transient. The actual status is being corrected to state that the reactor water cleanup pump tripped during the transient. The licensee has notified the NRC Resident Inspector. Notified R3DO (Stone).
ENS 50210	18 June 2014 19:45:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Loss of Hpci Room Cooling	At 1545 (EDT), while testing of the Emergency Service Water system (ST-8Q) was being performed at the James A. FitzPatrick Nuclear Power Plant (JAF), two of five unit coolers (66UC-22H and 66UC-22K) in the East Crescent were found with indicated flow of 0 gpm. The other three unit coolers in the East Crescent Area were found with sufficient flow. At least four unit coolers are required to support the functionality of the East Crescent Area Ventilation Subsystem (TRO 3.7.C). The East and West Crescent Area Ventilation Subsystems support the Operability of the Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) system by removing heat from the areas, in the event that ECCS and RCIC were used to mitigate the consequences of an accident. The West Crescent Area Ventilation Subsystem remained functional. The accident mitigating function of the division of ECCS and RCIC located in the West Crescent Area were unaffected by this condition. However, this condition could have prevented the function of one division of the ECCS, including the single train of High Pressure Coolant Injection (HPCI), located in the East Crescent. Therefore, this condition could have prevented fulfillment of the safety function of HPCI and it is being reported under 10 CFR 50.72(b)(3)(v)(D). As part of the testing, the throttle valves to the unit coolers (66UC-22H and 66UC-22K) were cycled and normal flow was restored. This condition no longer exists. The licensee is investigating the loss of flow to the "H" and "K" unit coolers and the restoration of flow by cycling the unit cooler supply throttle valves. The licensee will be notifying the NRC Resident Inspector. * * RETRACTION FROM DAVID CALLEN TO DANIEL MILLS AT 1506 EDT ON 8/13/2014 * * * FitzPatrick is retracting EN # 50210 made on June 18, 2014 at 2120 EDT. The plant was at 86% power at the time. The ENS notification was an 8-Hr non-emergency notification to 10 CFR 50.72(b)(3)(v)(D) when it was discovered that two of five unit coolers in the East Crescent (66UC-22H and 66UC-22K) were found with indicated flow of 0 gpm while testing. The other three unit coolers in the East Crescent (66UC-22B, 66UC-22D, 66UC-22F) were found with sufficient flow. At least four unit coolers are required to support the functionality of the East Crescent Area Ventilation subsystem (TRO 3.7.C). The East and West Crescent Area Ventilation subsystems support the Operability of the Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) system by removing heat from these areas in the event that ECCS and RCIC are used to mitigate the consequences of an accident. As part of testing, throttle valves to unit coolers 66UC-22H and 66UC-22K were cycled and normal flow was restored. The West Crescent Area Ventilation subsystem remained functional. The accident mitigating function of the division of the ECCS and RCIC located in the West Crescent Area were unaffected by this condition. Initial review of this condition determined that it could have prevented the function of one division of the ECCS, including the single train of High Pressure Coolant Injection (HPCI), located in the East Crescent. Therefore, this condition was initially reported under 10 CFR 50.72 (b)(3)(v)(D) as a condition that could have prevented fulfillment of the Safety function of HPCI. This EN# 50210 is being retracted based upon a subsequent engineering analysis that determined that there is reasonable assurance that the three unit coolers with sufficient flow (66UC-22B, 66UC-22D, and 66UC-22F) would have been capable of removing accident heat loads as a function of time to maintain East Crescent area temperatures at a value which ensures operability of supported equipment. The analysis considered unit cooler heat transfer capability at the modified design condition flow of 22 gpm for historically observed lake temperatures and for flow at tested conditions. Additional margin in flow at the tested condition provided increased heat removal capability and provided added assurance that accident heat load would have been removed. The East Crescent Area Ventilation subsystem was, therefore, functional with three unit coolers (functionality never was lost) and the supported ECCS remained Operable. The Operability determination for the condition has subsequently been revised based upon the engineering analysis, to state the condition was not immediately reportable per 10 CFR 50.72. The licensee has notified the NRC Resident Inspector Notified R1DO (Kennedy)
ENS 49928	19 March 2014 03:52:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Reactor Scram Due to Failure to Control Main Turbine Moisture Separator Level	At 2252 on 03/18/2014, the Unit 3 reactor automatically scrammed due to a turbine trip from a high Main Turbine moisture separator level. Initial indications show the level controller for 3B2 Moisture Separator failed to adequately maintain level. Additionally local manual control attempts failed to restore moisture separator level. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System Groups 2, 3, 6 and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. The reactor had been operating near 35% power during scheduled power ascension. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). NRC Resident Inspector has been notified.
ENS 49920	17 March 2014 10:14:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram Due to Steam Leak in Low Pressure Turbine Line	On 3/17/2014 at 0514 (CDT) the reactor was manually scrammed from approximately 41% core thermal power due to a steam leak in the turbine building. All control rods fully inserted and all systems actuated and operated as designed. All Main Steam Isolation Valves were manually shut. The Reactor Core Isolation Cooling System was manually initiated to assist in level control and pressure control. No safety relief valves actuated automatically. Manual cycling of safety relief valves and Reactor Core Isolation Cooling are being used to maintain reactor water level and pressure within normal bands. Group 2 and 3 RHR isolation signals were received; however no valve movement occurred since the affected valves are normally closed. 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 licensee informed the NRC Resident Inspector.
ENS 49888	10 March 2014 20:28:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Reactor Scram Due to Actuation of the Alternate Rod Insertion System	At 1628 EDT Nine Mile Point (NMP) Unit 2 experienced an actuation of the Alternate Rod Insertion (ARI) system which resulted in a reactor scram. Coincident with the scram, the Reactor Core Isolation Cooling (RCIC) system initiated. Prior to the event, maintenance personnel were working in the vicinity of a reactor vessel level instrumentation rack and may have agitated the common drain line of the transmitters. A prompt investigation is underway to investigate the incident. The actuation signal for the RCIC system was invalid because reactor vessel level did not reach level two and the actuation was not in response to actual plant conditions or parameters. The reactor scram is reportable in accordance with 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 event has been entered into the NMP corrective action program as CR-2014-001963. The NRC Resident Inspector has been notified. The licensee has notified the State of New York. The reactor is shutdown with all rods inserted. Decay heat is being rejected to the condenser and reactor water level is being maintained by condensate, feedwater, reactor water clean up, and control rod drive systems.
ENS 49407	4 October 2013 01:45:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Unit 2 High Pressure Coolant Injection (Hpci) Inoperable Due to Drain Line Leak	On October 3, 2013, at 2045 (CDT) hours, a defect (pinhole through-wall leak) was identified on the drain line for the LS 2-2365, HPCI TURBINE INLET DRAIN POT LEVEL SWITCH. The defect was identified during investigation of leakage near LS 2-2365. The LS 2-2365, HPCI TURBINE INLET DRAIN POT LEVEL SWITCH, is provided to detect a failure of the HPCI steam trap during standby line-up. The location of the defect, is in Class 2 Safety related piping. HPCI is a single train safety system and this notification is being made in accordance with 10CFR50.72(b)(3)(v)(D). The instrument isolations for LS 2-2365 have been close and the leak has been isolated. There is no increase to plant risk and RCIC (Reactor Core Isolation Cooling) is available. The licensee will inform the NRC Resident Inspector. * * RETRACTION ON 11/7/13 AT 1412 EST FROM JEFFERY JACOBSON TO DONG PARK * * * The purpose of this notification is to retract the ENS report made on October 4, 2013, at 0212 EDT (ENS Report # 49407). Upon further investigation the pinhole through-wall leak discovered in the Unit 2 HPCI room was in a weld at a 'Tee' downstream of the Unit 2 HPCI turbine inlet drain pot level switch (LS 2-2365) on drain line 2-2386B-1-B. The defect was characterized as a 1/16-inch rounded hole due to gas porosity (with no evidence of cracking). A subsequent evaluation performed by Quad Cities Station considering the defect size, location, and characterization confirmed the Unit 2 High Pressure Coolant Injection (HPCI) system would have performed its safety function when required. Based on this subsequent evaluation, ENS Report # 49407 is being retracted. Note: On October 3, 2013, at 1155 CDT the Unit 2 HPCI drain line leak was isolated and HPCI was declared operable. The licensee has notified the NRC Resident Inspector. Notified R3DO (Lipa).
ENS 49296	22 August 2013 11:55:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Reactor Protection Actuation (Scram)	On Thursday, August 22, 2013 at 0755 hours (EDT), with the reactor critical at approximately 98% core thermal power, and the mode switch in RUN, a manual reactor scram was inserted due to lowering reactor water level. The cause of the lowering reactor water level was due to the trip of all three Feedwater Pumps. The cause of the Feedwater Pump trip event is currently under investigation. Following the reactor scram, all control rods were verified to be fully inserted. All 4kV busses transferred to the Startup Transformer as designed. Following the scram the reactor water level lowered to +12 inches initiating the Primary Containment Isolation System (Group II, Reactor Building Isolation System (RBIS); and Group VI - Reactor Water Cleanup System) automatically as per design. Reactor water level lowered to -46 inches initiating Primary Containment Isolation System Group I - Main Steam Isolation Valves (MSIVs); Emergency Core Cooling Systems (ECCS) actuated which included automatic start and injection of the High Pressure Coolant Injection (HPCI) System and the Reactor Core Isolation Cooling (RCIC) System and an automatic start of the Emergency Diesel Generators as designed. Reactor water level was promptly restored to normal level. Currently a cooldown is in progress with reactor pressure is being maintained by the HPCI System operating in the pressure control mode and reactor water level is being maintained by the RCIC System. Reactor Water Clean-up System and normal reactor building ventilation have been restored. Off-site power is being supplied to the station by the Start-up Transformer (normal power supply for shutdown operations). This event had no impact on the health and/or safety of the public. The USNRC Senior Resident Inspector has been notified. This 4-hour notification is being made in accordance with 10 CFR 50.72 (b)(2)(iv)(A) and (B). The plant is transferring from decay heat removal to the torus to decay heat removal to the main condenser. Reactor pressure is 371 psig. Initial indications are that a main feedwater power supply breaker tripped.
ENS 49225	30 July 2013 19:32:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Reactor Scram Due to Turbine Generator Trip	Actuation of Reactor Protection System with reactor critical. Reactor Scram occurred at 1432 CDT on 7/30/2013 from 100% Power. The cause of the scram appears to be a Turbine Generator trip. 05-S-01-EP-2, 'Reactor Pressure Vessel Control,' 05-1-02-I-1, 'Reactor Scram Off Normal Event Procedure,' and 05-1-02-I-2, 'Turbine and Generator Trip Off Normal Event Procedure,' were entered to mitigate the transient. No Loss of Off-site Power occurred. No Emergency Core Cooling System or Diesel Generator initiation occurred. Reactor Core Isolation Cooling initiated and injected. The lowest reactor water level reached was -36 inches wide range (RCIC initiation set point is -41.6 inches wide range). Main Steam Isolation Valves remained open and no Safety Relief Valves actuated. Currently, Main Turbine Bypass valves are controlling reactor pressure to the Main Condenser and Condensate and Feedwater is controlling reactor water level in the normal band and removing decay heat. There are no challenges to Primary or Secondary Containment. The NRC Senior Resident Inspector was notified. * * * UPDATE FROM CHRIS ROBINSON TO PETE SNYDER AT 1841 EDT ON 7/30/13 * * * The first out recorder indicated that RPS actuation signal was due to high reactor pressure as a result of the turbine control valves going shut. Notified R4DO (Farnholtz).
ENS 49084	2 April 2013 13:46:00	10 CFR 50.73(a)(1), Submit an LER	6O-Day Optional Telephone Notification for an Invalid Specified System Actuation	This 60-day telephone notification is being made per the reporting requirements specified in 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to describe an invalid actuation signal affecting containment isolation valves in more than one system. On April 2, 2013, Nine Mile Point 2 (NMP2) received a Division 2 reactor building area high ambient temperature isolation signal when lifting a lead for trip unit E31-N638B while performing surveillance N2-IPS-LDS-Q010, Reactor Building General Area Temperature Instrumentation Channel Functional Test. The isolation signal provided a closure signal to two Reactor Core Isolation Cooling System (RCIC) valves, and three Residual Heat Removal (RHR) system containment isolation valves. As a result of the isolation signal one of the RCIC containment isolation valves, 2ICSMOV128 closed. The other four valves were already in their normal closed position. The RHR system valves are associated with the RHR Shutdown Cooling System and second RCIC isolation valve is used to warmup and place the RCIC system in standby following an isolated condition. All affected isolation valves responded as designed. As a result of 2ICSMOV128 closing the RCIC system was declared inoperable. Technical Specification 3.5.3, RCIC System, Condition A was entered. Action A.1 required verifying the High Pressure Core Spray System (HPCS) was operable immediately. Action A.2 requires restoring RCIC to operable within 14 days. After the instrumentation system was restored to normal, the RCIC system was subsequently restored to available later that day at 1205 (EDT) and operable at 1500 (EDT). The actuation signal was not valid because it resulted from maintenance activities when leads were lifted, and the trip unit had not been bypassed as required by the procedure. There were no isolation logic signals in response to actual plant conditions or parameters. This event was entered into the corrective action system as Condition Report (CR) 2013-002461. There were no actual safety consequences or impact on the health and safety of the public as a result of this event. The licensee notified the NRC Resident Inspector and the State.
ENS 48977	28 April 2013 02:24:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded	Technical Specification Required Plant Shutdown	This notification is being provided in accordance with 10CFR50.72(b)(2)(i), Plant Shutdown required by Technical Specifications, and 10CFR50.72(b)(3)(ii)A, Degraded or Unanalyzed Condition. At 2245 CDT on 04/27/13, LaSalle Unit 1 commenced a Technical Specification required plant shutdown, due to identification of pressure boundary leakage. At 2124 CDT on 04/27/13, a through-wall leak was identified in the body of 1E51-F076, Reactor Core Isolation Cooling system steam supply inboard isolation bypass warmup valve. This qualifies as pressure boundary leakage, which requires entry into Technical Specification 3.4.5, Reactor Coolant System Operational Leakage, Required Action C, to be in Mode 3, Hot Shutdown, by 0924 (CDT) on 04/28/13, and Mode 4, Cold Shutdown, by 0924 (CDT) on 04/29/13. This leakage is significantly less than 10 gpm and therefore does not meet the threshold for entry into the Emergency Action Plan. At the time of discovery, Unit 1 was in startup mode following a forced outage. A unit shutdown has been initiated. A repair plan is being prepared at this time, and the unit will remain in Cold Shutdown until repairs are complete. The leak is located inside the primary containment and was visually identified during a containment walk-down. The licensee has notified the NRC Resident Inspector.
ENS 48969	26 April 2013 01:19:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Reactor Scram Following Trip of Circ Water Pumps	This report is being made pursuant to 10CFR50.72(b)(2)(iv)(B), RPS Actuation (scram). At 2019 CDT on April 25, 2013, LaSalle Unit 2 was manually scrammed due to a loss of Condenser Circulating Water. The Unit was manually scrammed after the condenser circulating water pumps tripped due to high level in the turbine building condenser pit. The high level in the condenser pit was caused by a leak on the upper manway of the condenser water box during a maintenance activity. MSIV's were isolated due to loss of heat sink. The safety relief valves were used in pressure control mode. Current plant status: reactor level is stable and reactor pressure is stable. The condenser water box manway leak has been isolated. The plant will remain in hot shutdown pending investigation and repairs. Reactor Core Isolation Cooling (RCIC) is being used in the pressure control mode. The licensee has notified the NRC Resident Inspector.
ENS 48939	17 April 2013 20:11:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Notification of Unusual Event Declared Due to Loss of Offsite Power from a Lightning Strike	LaSalle Unit 1 and LaSalle Unit 2 have both experienced an automatic reactor scram, in conjunction with a loss of offsite power. This was caused by an apparent lightning strike in the main 345kV/138kV switchyard during a thunderstorm. 138kV line 0112 has been inspected in the field, and heavy damage has been noted on the insulators on two of the three phases on a line lightning arrestor line side. The plant systems have all responded as expected. All five diesel generators started, and have loaded on to their respective buses as designed. All rods went full in on both units during the respective scrams. HPCS (High Pressure Core Spray) system was started on each unit and automatically aligned for injection for initial level control. The MSIVs (Main Steam Isolation Valves) are shut on both units with decay heat being removed via the safety relief valves. Suppression pool cooling is in progress. The licensee will notify the NRC Resident Inspector and has notified the State. Notified DHS, FEMA, USDA, HHS, DOE, NICC, EPA, and Nuclear SSA via email. * * UPDATE FROM DON PUCKETT TO VINCE KLCO AT 2113 EDT ON 4/17/2013 * * * In addition to information (previously provided), LaSalle Unit 2 received a high drywell pressure signal (1.77 psig) due to loss of containment cooling from the loss of power. At the time of this high drywell pressure signal, high pressure core spray pump and 2B residual heat removal (RHR) pump was already in operation, the low pressure core spray system and 2A residual heat removal system was secured and (placed) in pull to lock. When the signal was satisfied the ECCS (Emergency Core Cooling Systems) signal was processed but only the 2C RHR pump would have started. In this case, the 2C RHR pump tripped when the signal was received. There is no evidence of reactor coolant leakage. There was no additional ECCS systems discharging into the RCS (Reactor Coolant System). As (initially stated), level was controlled using High Pressure Core Spray and level control is now being maintained using the Reactor Core Isolation Cooling (RCIC) systems. The 2C RHR pump trip is under investigation. Due to the initial loss of offsite power for both Unit 1 and Unit 2 reported at 1511 (CDT), multiple containment isolation valves isolated and closed as expected. Once initial containment isolations were verified, two Unit 2 primary containment vent and purge valves were opened to vent the Unit 2 containment. Once Unit Two containment pressure reached 1.77 (psig), these two vent valves isolated as expected. Due to the loss of offsite power, the Station Vent Stack Wide Range Gas Monitor (WRGM) and the Standby Gas Treatment Wide Range Gas Monitor (VGWRGM) also lost power. Manual sampling has been implemented and power is restored to the VGWRGM, however the VGWRGM has not been declared operable yet. Normal radiation levels have been reported from the manual sampling. (This is being reported in accordance with 10CFR50.72(b)(3)(xiii).) The licensee notified the NRC Resident Inspector and the State of Illinois. Notified the R3 IRC, NRR EO(Skeen), IRD MOC (Grant). * * UPDATE AT 0057 EDT ON 04/18/13 FROM MIKE LAWRENCE TO S. SANDIN * * * After the Unit 2 primary containment vent and purge system isolated on the Unit 2 containment High Pressure signal, Venting of the Unit 1 primary containment was commenced. At 2005 CDT, Unit 1 primary containment pressure reached the Group 2 primary containment isolation system setpoint (1.77 PSIG) causing the primary containment vent and purge valves being used to vent the Unit 1 containment to isolate. Unit 1 primary containment venting was being performed through the Standby Gas Treatment system which is a filtered system. In addition to the primary containment isolation signal on high drywell pressure, an ECCS initiation on high drywell pressure also occurred. The ECCS signal resulted in an auto start of the 1C RHR system. The 1B RHR system was already running in suppression pool cooling mode. 1A RHR and LPCS had been secured to prevent overloading the common diesel generator for division 1. The common diesel generator supplies both Unit 1 and Unit 2 division 1 ESF busses. The licensee informed the NRC Resident Inspector. Notified NRR EO (Skeen), IRD MOC (Grant) and R3IRC (Louden). * * UPDATE AT 0947 EDT ON 04/18/13 FROM JUSTIN FREEMAN TO PETE SNYDER * * * LaSalle has terminated the unusual event which was initiated at 1511 on 4/17/13 and reported under EN 48939. This unusual event has been terminated based on meeting the following established criteria. This report is being made in accordance with 10CFR50.72.(c)(1)(iii). 1) Off-site power has been restored to all ESF busses 2) Fuel Pool Cooling has been restored on both units 3) Primary Containment Chillers have been restored on both units 4) Drywell pressure is less than ECCS initiation setpoint 5) ECCS signals cleared to allow diesels to be placed in stand by Recovery of remaining plant systems will be managed through the Outage Control Center (OCC)." The licensee informed the NRC Resident Inspector. Notified R3DO (Orth), NRR EO (Chernoff), IRD (Grant), DHS, FEMA, USDA, HHS, DOE, NICC, EPA, and Nuclear SSA via email. * * UPDATE AT 1711 EDT ON 4/21/2013 FROM GREG LECHTENBERG TO MARK ABRAMOVITZ * * * In addition to the 10 CFR 50.72 Sections initially identified, the Loss of Offsite Power was also reportable under 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of systems needed mitigate the consequences of an accident. This event is considered a safety system functional failure for both Units 1 and 2. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Orth).
ENS 48904	11 February 2013 11:13:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Initiation of Reactor Core Isolation Cooling System	On February 11, 2013, at 0613 hours (CDT), the Reactor Core Isolation Cooling (RCIC) system was manually started during a planned Unit 3 reactor shutdown. A Reactor Feedwater recirculation piping separation resulted in the loss of condenser vacuum and subsequent unavailability of the Main Turbine Bypass Valves. The RCIC system was manually started at 9.2" of condenser vacuum in order to control reactor water level in anticipation of loss of Reactor Feedwater Pumps (RFPs) which occurs at 7" of condenser vacuum. Safety Relief Valves (SRVs) were manually operated to maintain reactor pressure. The reactor water level was controlled in the normal band by RCIC, and Reactor Pressure was controlled with a combination of Reactor Core Isolation Cooling (RCIC) system and SRV manual operation. All systems operated as designed and Reactor water level was maintained in the prescribed band. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. RCIC operation was secured at 1449 (CDT) on 2/11/2013. This event is reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A). During a review of operating logs it was identified that this event met reporting requirements and had not been reported. Therefore, this report does not comply with the 8 hour requirement. This condition has been entered into the corrective action program. Additionally, an LER is required within 60 days per 10CFR50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 48897	8 April 2013 15:15:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Hpci Declared Inoperable During Surveillance Testing	On April 8, 2013 at 0908 (EDT), the High Pressure Coolant Injection System (HPCI) was declared inoperable as part of planned Controls Functional Testing. At 1115 (EDT), during the performance of scheduled testing, an initiation signal for the HPCI system was provided and the HPCI Auxiliary Oil Pump failed to start as expected. The HPCI Auxiliary Oil Pump provides the motive force to open the HPCI Turbine Stop and Governor valves during system startup. The inability of the HPCI Turbine Stop and Governor valves to open prevents the HPCI system from fulfilling its design safety function. The HPCI system will remain inoperable until the cause of the failure has been corrected. All other Emergency Core Cooling Systems and the Reactor Core Isolation Cooling (RCIC) system remain operable. The unit remains at 100% power. The station has initiated an Event Response Team to identify and correct the cause of the failure. No personnel injuries resulted from the event. The NRC Resident Inspector and Lower Alloways Creek Township will be notified.
ENS 48782	25 February 2013 19:13:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Automatic Scram Due to a Turbine Trip from a Loss of Condenser Vacuum	At 1313 (CST) on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 (CST) with positive indication of isolation, and isolation valves deactivated at time 1514 (CST) to satisfy TS LCO 3.6.1.3 required actions. This event is reportable within 4 hours per 10CFR50.72(b)(2)( iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). At 1415 (CST), Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 (CST), water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2.
ENS 48736	9 February 2013 03:00:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration	Unusual Event Declared Due to Loss of Offsite Power	Pilgrim Station scrammed on a loss of offsite power. All systems performed as designed. Groups I, II, VI went to completion. Reactor Core Isolation Cooling (RCIC) is injecting to the vessel controlling level. High Pressure Coolant Injection is in pressure control and slowly cooling down. Offsite power was lost multiple times. The Startup Transformer has been declared inoperable. The Unusual Event was declared under EAL SU 1.1 based on loss of offsite power greater than 15 minutes (at 2200 EST). The licensee originally experienced an automatic reactor scram at 2117 EST due to a load reject with a turbine trip/reactor scram due to loss of power. Offsite power availability has been fluctuating in and out to the site. The licensee states that all systems are functioning as required. All rods fully inserted and the reactor is stable in Mode 3. Both Emergency Diesel Generators are providing power to the safety related buses. The loss of offsite power is believed to be weather related. The licensee has notified the State and local authorities and the NRC Resident Inspector. Notified DHS SWO, FEMA, USDA, HHS, DOE, DHS NICC, EPA, and NuclearSSA via email. * * UPDATE FROM PAUL GALLANT TO VINCE KLCO AT 2/10/13 AT 1108 EST* * * Pilgrim terminated the Unusual Event and has transitioned to recovery effective at 10:55 AM on 02/10/2013. Offsite power has been restored to safety-related and non-safety-related electrical buses through the station Startup Transformer via a single 345 KV line. The other two offsite power sources remain out of service. The emergency diesel generators have been secured and are in standby. Residual heat removal is in shutdown cooling mode maintaining the reactor in cold shutdown. Fuel Pool Cooling is in service with fuel pool coolant temperatures trending down. The licensee notified State, local authorities and the NRC Resident Inspector. Notified R1 RA (Dean), R1DO (Powell), NRR DIR(Leeds) NRR EO (Evans) and NSIR IRD (Marshall). Notified DHS SWO, FEMA, USDA, HHS, DOE, DHS NICC, EPA, and NuclearSSA via email.
ENS 48696	23 January 2013 20:16:00	10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Concurrent Loss of High Pressure Reactor Makeup Systems Capability	On 1/23/2013 at 1516 (EST), Nine Mile Point 2 (NMP2) had a failure of a Reactor Building General Area temperature trip unit occur resulting in the closure of an isolation valve on the Reactor Core Isolation Cooling (RCIC) system steam supply line. Concurrent with this failure, the High Pressure Core Spray (HPCS) system was inoperable for planned surveillance testing. With both the RCIC and HPCS systems inoperable, NMP2 entered a Technical Specification Required Action to be in Mode 3 within 12 hours. At 1550, the HPCS system was restored to OPERABLE. Based on the concurrent loss of the high pressure reactor makeup capability of these two systems, it was determined that the condition is reportable under section 50.72(b)(3)(v) as the following safety functions were impacted: (A) Shutdown the reactor and maintain it in a safe shutdown condition; and (D) Mitigate the consequences of an accident. NMP2 remains in a stable condition at rated power. The offsite grid is stable with no restrictions or warnings in effect. The licensee notified the NRC Resident Inspector.
ENS 48688	22 January 2013 08:32:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Reactor Protection System Actuation	On January 22, 2013, at approximately 0332 hours (EDT), an automatic Reactor Protection System (RPS) actuation occurred at the Perry Nuclear Power Plant, Unit 1. At the time of the event, the plant was in Mode 1 at 100% power. All control rods are inserted into the reactor core and the plant is currently stable in Mode 3 (Hot Shutdown) with reactor pressure and level being maintained in the normal shutdown range. The RPS actuation was initiated by a low reactor water level (Level 3 - 178") signal. In response to the RPS actuation and subsequent reactor Level 2 (130") signal, the High Pressure Core Spray (HPCS) system and Reactor Core Isolation Cooling (RCIC) system both actuated and injected to maintain reactor coolant level. The reactor level is currently being maintained in its normal band by the feedwater system and decay heat is being removed by (turbine bypass valves to) the condenser (both HPCS and RCIC have been returned to standby). The plant is in a normal electrical line-up with all three Emergency Diesel Generators operable and available, if needed. The Containment Isolation Valves (responded to the Level 2 and 3) isolation signals as designed. The cause of the RPS actuation is under investigation. The NRC Resident Inspector has been notified.
ENS 48623	22 December 2012 17:52:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram Due to Loss of Power to the Reactor Protection System	On 12/22/2012 at 1152 CST, the Unit 2 reactor automatically scrammed due to actuation of the Reactor Protection System (RPS) from loss of power to RPS. At 1134 CST, the D 4kV Shutdown Board unexpectedly de-energized during performance of post-maintenance testing for the 3D Diesel Generator paralleling circuitry, resulting in loss of power to the 2B RPS subsystem. Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations were received along with automatic initiation of A, B, and C Standby Gas Treatment subsystems and A Control Room Emergency Ventilation subsystem due to loss of power to the 2B RPS subsystem. While attempting to reenergize the 2B RPS subsystem, the 2A RPS subsystem was inadvertently de-energized resulting in Unit 2 reactor automatic scram. All affected safety systems responded as expected for the loss of RPS and reactor scram. Due to the loss of RPS, the Main Steam Isolation Valves (MSIVs) closed. Reactor pressure did not rise to the automatic initiation set point for Safety Relief Valve (SRV) actuation. Reactor Core Isolation Cooling System (RCIC) and High Pressure Coolant Injection System (HPCI) reactor water level initiation set point of -45" was reached and RCIC and HPCI automatically initiated as designed to restore water level above the initiation set point. Both Recirculation Pumps also tripped on reactor water level of -45". Reactor pressure control was established by manually operating one SRV and water level control established with RCIC. HPCI was returned to standby readiness. The scram was reset, MSIVs were opened, and the Main Condenser was established as a heat sink. The scram event from critical is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B). It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector was notified. The 2A and 2B RPS subsystems were returned to service. The electrical grid is stable and supplying shutdown loads on Unit 2. Unit 1 and Unit 3 were unaffected and continue to operate at 100% power.
ENS 48570	12 October 2012 14:12:00	10 CFR 50.73(a)(1), Submit an LER	Temperature Switch Failure Causes Division I Isolation Signal	This 60-day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation signal affecting containment isolation valves in more than one system. On October 12, 2012, Nine Mile Point Unit 2 (NMP2) received a Division I primary containment isolation signal which resulted in the closure of Group 5, 6, and 10 primary containment isolation valves (PCIVs) in the following systems: Group 5 PCIVs: Residual Heat Removal System (RHS); Shutdown Cooling (SDC); Group 6 PCIVs: Reactor Water Cleanup System (WCS) supply outside isolation valve; and Group 10 PCIVs: Reactor Core Isolation Cooling (RCIC) System. All affected PCIVs responded as designed. The Division I isolation signal was generated due to the failure of a temperature switch unit. The Division I and II temperature switch units were both reading within limits when the Division I unit failed. Since the isolation signal was not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation, the isolation signal was determined to be invalid. The event was entered into the corrective action program as Condition Report CR-2012-009380. There were no safety consequences and no impact to the health and safety of the public as a result of this event. The licensee notified the New York State Public Service Commission and the NRC Resident Inspector.
ENS 48479	5 November 2012 02:53:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Automatic Reactor Scram from Full Power Following Turbine Trip	The reactor was scrammed on a valid reactor protection system activation caused by a main turbine trip. The cause of the main turbine trip is under investigation. All control rods fully inserted. All isolations and initiations occurred as designed. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated as expected. RCIC injected into the reactor coolant system, HPCI did not, as expected. This scram was characterized as uncomplicated and the reactor is stable in Mode 3. The plant is in a normal post shutdown electrical lineup. All systems functioned as required. The NRC Resident Inspector has been notified.
ENS 48341	25 September 2012 15:44:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram During Maintenance on 4160V Bus 12 Ammeter	During maintenance on 4160V Bus 12 ammeter, a Bus 12 lockout occurred. The station power was from 1R Reserve transformer for work on the 2R Auxiliary transformer. Net effect was Bus 12 locked out, removing power from 12 Reactor Feed Pump and 12 Reactor Recirculation pump. Reactor level lowered to +23 inches then began to rise. With both Main Feed Reg Valves in AUTO, the level transient reached +48 inches, the Reactor Water Level Hi Hi setpoint. The Main Turbine and 11 Reactor Feed Pump tripped as designed, and a Reactor SCRAM occurred. Reactor water level began to drop, and C.4.A Abnormal Procedure for SCRAM was used to restart 11 Reactor Feed Pump and recover water level. Minimum water level reached was -26 inches. Reactor Low Level SCRAM signal and Group 2 Primary Containment isolation occurred at +9 inches as designed, No Safety Relief valves lifted during this transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) did not receive an initiation signal due to not reaching their setpoints. There were no Emergency Core Cooling Systems initiation setpoints reached. Prior to the event, both divisions of Standby Liquid Control were inoperable as part of planned maintenance. All control rods fully inserted. Decay heat is being removed through the turbine bypass to the main condenser. The plant is in a normal shutdown electrical lineup and stable in Mode 3. The licensee has notified the NRC Resident Inspector and will notify the State and local governments.
ENS 48270	3 September 2012 06:25:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Hpci Inoperable Due to Erroneous Indication on Flow Indicating Controller	At 0225 EDT on September 3, 2012, with the James A. Fitzpatrick Nuclear Power Plant (JAF) operating at 93% reactor power, High Pressure Coolant Injection (HPCI) was declared inoperable due to abnormal indication on the HPCI Flow Indicating Controller (FIC). The FIC was found to be indicating a HPCI System flow rate of 700 gpm while the system was in the standby lineup. Under these conditions, the capability of the system to achieve the required flow rate cannot be assured. This failure meets NRC 8 hour reporting criterion 10CFR50.72(b)(3)(v)(D). Reactor Core Isolation Cooling (RCIC) and other Emergency Core Cooling Systems (ECCS) remain operable. The NRC Resident Inspector has been notified. * * UPDATE AT 1418 EDT ON 9/4/12 FROM DeFILLIPPO TO HUFFMAN * * * The improper HPCI flow indication was determined to be due to minor air intrusion following restoration of the system after maintenance. The flow transmitter for the HPCI system was repeatedly vented with no air observed. The HPCI system has been restored to a normal standby line-up and is OPERABLE as of 9/4/2012 at 1415 EDT. The NRC Resident Inspector has been notified. R1DO (Conte) notified.
ENS 48258	30 August 2012 16:15:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Hpci Inoperable Due to Failed Pressure Control Valve	At 1215 EDT on August 30, 2012, with the James A. FitzPatrick Nuclear Power Plant (JAF) operating at 95% reactor power, High Pressure Coolant Injection (HPCI) was declared inoperable due to the failure of a pressure control valve on the HPCI oil cooling system. The failure of this pressure control valve results in a safety-valve lifting and releasing approximately 75 gallons per minute to the reactor building equipment drain tank. There was no release to the environment. This failure meets NRC 8 hour reporting criterion 10CFR50.72(b)(3)(v)(D). Reactor Core Isolation Cooling (RCIC) and other Emergency Core Cooling System (ECCS) systems remain operable. The NRC Resident Inspector has been notified.

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