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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56802 | 18 October 2023 15:16:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Auxiliary Feedwater System | The following information was provided by the licensee via email: On October 18, 2023, at 1116 (EDT), with Unit 1 in Mode 5, an automatic actuation of the 1A auxiliary feedwater motor driven pump occurred when an incorrect action resulted in an automatic start signal. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. Feedwater is not needed for plant conditions, and the 1A auxiliary feedwater pump did not feed the steam generators. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 56451 | 2 April 2023 07:52:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Actuation of Auxiliary Feedwater Motor Driven Pumps | The following information was provided by the licensee via email: On April 2, 2023, at 0341 EDT, with Unit 2 in Mode 3 and the 2B main feedwater pump feeding the steam generators, the 2A main feedwater pump recirculation valve, 2CF-76, failed. Further, observation of the operating 2B main feedwater pump recirculation valve, 2CF-81, called into question its functionality. At 0352 EDT, operations manually started the auxiliary feedwater motor driven pumps to feed the stream generators to allow maintenance on the main feedwater system. The auxiliary feedwater motor driven pumps started as designed. Flow to the steam generator was not adversely impacted during this sequence. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the auxiliary feedwater system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 55748 | 18 February 2022 09:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and System Actuation | The following information was provided by the licensee via telephone and email: On 2/18/2022, McGuire Nuclear Station Unit 2 experienced a turbine runback to 55 percent power. Based on concerns with unit stability, the reactor was manually tripped at 0459 (EST). All Auxiliary Feedwater pumps started on low steam generator level as required. The reactor trip was uncomplicated with all systems responding normally post trip. A feedwater isolation occurred as designed. Unit 1 was not affected. Due to the Reactor Protection System actuation while critical, actuation of the Turbine Driven Auxiliary Feedwater Pump and Motor Driven Auxiliary Feedwater pumps along with the Feedwater Isolation, this event is being reported as a four hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour nonemergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods fully inserted. Decay heat is being removed via the condenser and normal feedwater. Unit 2 is in a normal shutdown electrical lineup. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 55458 | 13 September 2021 04:11:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation | At 0011 EDT, with Unit 2 in Mode 5 (Cold Shutdown), actuations of the 2B Diesel Generator (DG) and the 2B Motor Driven Auxiliary Feedwater (AFW) Pump occurred during Engineered Safety Features Actuation Periodic Testing while resetting the 2B DG Load Sequencer. The 2B DG was running unloaded following test actuation, and during realignment from the test, a blackout condition was experienced when the breaker opened supplying the 4160 Volt Essential Power System 2ETB from the Standby Auxiliary Power Transformer SATB. Sequencer actuation closed the emergency breaker to 2ETB and loaded the 2B Motor Driven AFW Pump onto the bus. Steam supply valves to the Turbine Driven AFW Pump were open from the previous test configuration. This event is being reported in accordance with 10CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the 2B DG and the 2B Motor Driven AFW Pump. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Auxiliary Feedwater | |
| ENS 54047 | 3 May 2019 19:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Over Temperature Delta Temperature | At 1554 EDT on 5/3/19, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped on Over Temperature Delta Temperature following a pressure transient in the Reactor Coolant System. The trip was uncomplicated with all systems responding normally post trip. Operations manually started the motor driven auxiliary feedwater pumps and has stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected. Due to Reactor Protection System actuation while critical and actuation of the motor driven auxiliary feedwater pumps, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Unit 1 is in a normal electrical lineup. Prior to the automatic trip, the backup pressurizer heaters were in service as is normal during power ascension. The pressure transient started when the backup heaters were in the process of being removed from service. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Reactor Protection System Auxiliary Feedwater | |
| ENS 53217 | 16 February 2018 15:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Solid State Protection System Testing | At 1014 (EST) hours on 2/16/18, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped when the Reactor Trip Breakers opened during Train B Solid State Protection System (SSPS) testing. The trip was uncomplicated with all systems responding normally post-trip. Operations manually started the motor driven auxiliary feedwater pumps. The turbine driven auxiliary feedwater pump (TDCAP) auto-started on low steam generator level. A Feedwater Isolation occurred as designed due to the Reactor Trip and Lo Tave condition. Operations stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected. Due to the Reactor Protection System actuation while critical, actuation of the TDCAP and motor driven auxiliary feedwater pumps along with the Feedwater Isolation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 51474 | 7 October 2015 10:55:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Afw Actuation During Refueling Outage Testing | On October 7, 2015, with McGuire Nuclear Station Unit 2 in Mode 4, operators were testing the main turbine and main feedwater pump turbines, 2A safety injection (SI) train trip functions. At the time of the test, the 2A and 2B Auxiliary Feedwater (AFW) pumps were in operation to provide make-up to the steam generators. The main feedwater system was not in service. During realignment activities from the 2A Sl test, the 2A AFW train actuation signal was unblocked when the '2A AFW auto start defeat' switch was returned to 'reset.' This caused the 2A AFW train control valves to fully open, and the associated steam generator sampling and blowdown valves to close. The actuation occurred as designed and there was no adverse impact to the Unit. Public health and safety were not impacted by this event. Based on a review of the Event Reporting Guidelines and the plant licensing basis, this event was initially determined to be an invalid actuation. However, after further review and discussions with the NRC, Duke Energy concluded the event should be reported as an 8-hour non-emergency in accordance with 10 CFR 50.72 (b)(3)(iv)(A) as a valid actuation of the Auxiliary Feedwater System. A Nuclear Condition Report was initiated for the late notification. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater Main Turbine | |
| ENS 49538 | 14 November 2013 18:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Indication of Dropped Control Rods | On November 14, 2013, at 1313 Eastern Standard Time, Unit 1 was manually tripped from 100% power due to indications of (four) dropped control rods. This manual reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The cause of the dropped rods is not confirmed at this time, but may be related to maintenance in a Rod Control Power Cabinet ongoing at the time of the event. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip and all plant systems operated as designed. The Auxiliary Feedwater (AFW) system (1A and 1B motor-driven pumps) was manually started for steam generator level control following reactor trip. The start of the AFW system is reportable per 10 CFR 50.72 (b)(3)(iv)(A) for a valid system actuation. Decay heat is being removed via the steam generators (via steam dumps to the main condenser). This event does not impact public health and safety. Unit 2 was not affected by this event. The licensee notified the NRC Resident Inspector.
A subsequent licensee evaluation determined that there were ten dropped control rods. Notified the R1DO (Desai). | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 48775 | 21 February 2013 14:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip on Loss of Both Main Feed Pumps | At 0957 EST on 02/21/13, Unit 1 Reactor automatically tripped from 100% power, due to a turbine trip. The turbine trip was caused by a loss of both main feedwater (MFW) pumps. The 1A motor driven auxiliary feedwater (AFW) pump auto started to feed the "A" and "B" Steam Generators (S/G). The 1B motor driven AFW pump was unavailable due to planned maintenance, so the turbine driven AFW pump was manually started to feed the "C" and "D" S/Gs. The reactor trip was uncomplicated. All control rods fully inserted. Decay heat removal is to the main condenser via the turbine bypass valves. There was no primary to secondary leakage. Electrical buses are being supplied via offsite power. Steam generator levels are being returned to normal and MFW has been reset and is available. All other plant systems functioned as designed during and after the reactor trip. There is no impact on Unit 2. There is no impact on the health and safety of the public. The loss of the MFW pumps is still under investigation. The licensee has informed the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 48549 | 2 December 2012 04:17:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Amsac Actuated at Lower than Expected Turbine Inlet Pressure | The Anticipated Transient Without Scram Mitigation System Actuation Circuitry (AMSAC) caused a Unit 2 turbine trip with reactor power at 31%. The 2A and 2B Auxiliary Feedwater pumps automatically started. The 2B Nuclear Service Water pump started as a result of 2B Auxiliary Feedwater pump automatically starting. The Main Feedwater Regulating valves and the Main Feedwater Bypass valves were in the correct position for corresponding power level and turbine inlet pressure, but AMSAC actuated earlier than design (290 psig vs. 360 psig). The licensee has notified the NRC Resident Inspector. | Feedwater Service water Auxiliary Feedwater | 05000370/LER-2012-002 |
| ENS 46709 | 30 March 2011 04:10:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Control Rod Malfunction During Rod Testing Results in Operators Manually Opening Trip Breakers | Rod L-13 did not function as expected during control rod movement test. This rod is in Shutdown Bank C. When withdrawing this bank, rod L-13 did not withdraw and when the bank was manually inserted, rod L-13 began to withdraw. The (operating) crew went to Enclosure 13.2 of the procedure to deal with the misaligned rods. This enclosure has procedural guidance to open the reactor trip breakers, if desired. The reactor trip breakers were opened and all 211 rods are fully inserted. The reactor was not critical. This activity was performed twice (at the request of reactor engineering). The licensee will remain in Mode 5 (Cold Shutdown) until troubleshooting and repair is completed. The licensee will be notifying the NRC Resident Inspector.
This notification pertains to Event Number 46709. Based on further investigation, this event is being retracted. The event described in Event Number 46709 involved a control rod malfunction on Unit 2 while in Mode 5, during RCCA movement testing. Specifically, Control Rod 'L-13' in Shutdown Bank 'c' did not move with the bank when the bank was withdrawn from the bottom of the core. When the bank was reinserted to the bottom of the core, L-13 was observed to be 12 steps withdrawn. This condition was corrected by opening the reactor trip breakers which placed L-13 at the bottom of the core. The subsequent troubleshooting plan involved further manipulation of Shutdown Bank 'C' with additional instrumentation on the rod control cabinets. The same anomaly occurred and the reactor trip breakers were again opened. This event (both openings of the reactor trip breakers) was reported to the NRC on 3/30/11 as a valid RPS actuation (8-hour report; 10 CFR 50.72(b)(3)(iv)(A)). The event in question did not result in any consequences, given that the plant was in Mode 5 and not critical. NUREG-1022, Revision 2 states that actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event (e.g., at the discretion of the licensee as part of a preplanned procedure). Notwithstanding the issue of whether opening the reactor trip breakers was to mitigate the consequences of an event, NUREG-1022 cites one valid example of actuations that need not be reported, namely if the actuation was 'at the discretion of the licensee as part of a preplanned procedure'. The purpose of the test being conducted was to identify issues with the control rod system. The malfunction that occurred is one of a host of possible issues that could reasonably be expected to occur. Although the test personnel did not go into the test expecting the need to open the reactor trip breakers, the malfunction that occurred resulted in a desire to open the reactor trip breakers in order to restore the plant to the desired configuration. This action was a choice as allowed by the test procedure, and the personnel involved were aware of the result of the action before it occurred. Therefore, the event constituted a 'pre-planned sequence during testing', and was 'at the discretion of the licensee as part of a preplanned procedure.' Based upon the above considerations, the event does not meet the aforementioned criteria for an 8-hour report, and Event Number 46709 is therefore retracted. The licensee has notified the NRC Resident Inspector of this update. Notified the R2DO (Seymour). | Control Rod | |
| ENS 46559 | 20 January 2011 20:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Both Main Feed Pumps During Shutdown | Unit 1 performed a manual reactor trip (from 28% power) due to the trip of the 1B main feedwater pump trip concurrent with the 1A main feedwater pump having been previously tripped per the shutdown procedure. Reactor power was below the setpoint for an automatic reactor trip from a turbine trip. All rods fully inserted, heat removal is from auxiliary feedwater and condenser steam dumps, and the electrical system is in a normal alignment. The plant has stabilized at normal RCS pressure and temperature. Both Units 1 and 2 are being shutdown as per Tech Spec 3.0.3 associated with an inoperable Nuclear Service Water System. Unit 2 is at 40% and decreasing power due to the Tech Spec action and was not effected by the Unit 1 reactor trip. Both units will be placed in Mode 5. The licensee has notified the NRC Resident Inspector. | Feedwater Service water Auxiliary Feedwater | |
| ENS 46003 | 12 June 2010 10:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to a Control Rod Drop Indication | McGuire Unit 1 was operating at 44% power due to a previously dropped control rod. Indication was received of a second control rod drop and the reactor was manually tripped in accordance with abnormal operating procedure guidance. All control rods fully inserted. The auxiliary feedwater system was manually started due to an approaching autostart setpoint. The unit is stable in Mode 3 at normal operating temperature and pressure. Normal containment air release remains in progress. The steam generators are being fed through the auxiliary feedwater system and will transition to the normal feedwater system. Decay heat removal is to the condenser through the steam dumps. There was no impact on Unit 2. The licensee is investigating the cause of the control rod drop indication. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 44869 | 23 February 2009 09:06:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Inadvertent Actuation of the Turbine Driven Auxiliary Feedwater Pump | The Turbine Driven Auxiliary/Emergency Feedwater pump auto started when power was lost to the steam admittance solenoid valves. Additional feedwater to the steam generators caused an increase in power; in response, turbine load was manually decreased. At no time were any power limits exceeded. Power level is currently at 87% and stable. Maximum power, as determined by core delta temperatures, did not exceed 100%. The licensee has notified the NRC Resident Inspector.
Event Report 44869 Retraction: On February 23, 2009, McGuire Nuclear Station notified the NRC of an inadvertent actuation of the Unit 1 Turbine Driven Auxiliary/Emergency Feedwater (TDAFW) Pump (reference Event Report 44869). This event occurred when power was removed to the steam admittance solenoid valves for the Unit 1 TDAFW Pump prior to isolating the steam supply to the pump. Removing power to the steam admittance valves caused them to fail open, starting the pump. As per the requirements of 10 CFR 50.72(b)(3)(iv)(A), this event was reported as a valid actuation of the Unit 1 TDAFW Pump. Further evaluation of the event concluded that the actuation of the Unit 1 TDAFW Pump was not the result of a valid signal initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function provided by the pump. In addition, the actuation of the Unit 1 TDAFW Pump was not the result of an intentional manual initiation. Therefore, actuation of the Unit 1 TDAFW Pump represented an invalid actuation which is not reportable per the requirements of 10 CFR 50.72(b)(3)(iv)(A). Therefore, McGuire is retracting Event Report 44869. Note that McGuire will report this event in accordance with the requirements of 10 CFR 50.73. The licensee has notified the Resident Inspector. Notified R2DO (Musser). | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 44624 | 3 November 2008 07:07:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manually Opened Reactor Trip Circuit Breakers to Insert Control Bank "B | Manually opened reactor trip breakers in Mode 5 to insert control bank 'B' due to blown fuse in rod control cabinet. Licensee was moving control rod bank "B" following I & C work, and the control rod bank failed to move as expected. All other control rod banks were inserted into the core at the time of the event. EDG's and offsite power sources are OPERABLE, and there is no increase in plant risk. The licensee will inform the NRC Resident Inspector. * * * RETRACTION ON 12/31/08 AT 1326 FROM RICK ABBOTT TO PETE SNYDER * * * Regarding the NRC Event Number 44624 conveyed November 3, 2008, McGuire Nuclear Station has determined that manually opening the reactor trip breakers was not reportable and hereby retracts this notification. Upon further consideration it was determined that manually opening the reactor trip breakers was a conservative decision to fully insert control rods based on the failure mechanism causing a single rod to drop to the fully inserted position. Manual actuation of the reactor trip breakers was not required by abnormal procedures and was performed only after consultation between operations, engineering and senior station management agreed that this was the preferred option. Therefore, the decision to manually open the reactor trip breakers is considered to be a preplanned actuation of the reactor protection system and is not reportable. The licensee informed the NRC Resident Inspector. Notified R2DO (M. Lesser). | Reactor Protection System Control Rod | |
| ENS 44618 | 31 October 2008 16:52:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Rod Control Malfunction | Initial conditions: Low Power Physics Testing after outage. Performing Dynamic Rod Worth measurements. Single dropped rod (K-2) on Control Bank B occurred after movement disagreement between rod groups. Entered AP-14, 'Rod Control Malfunction'. Performed a manual reactor trip and manual start of Auxiliary Feedwater as directed by Normal Operating Procedures for unit shutdown. Feedwater isolation occurred due to Low T-Ave concurrent with reactor trip. All control rods were fully inserted. Licensee has notified NRC Resident Inspector. | Feedwater Auxiliary Feedwater Control Rod | |
| ENS 44318 | 26 June 2008 21:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Loss of 1B Reactor Coolant Pump | Loss of 1B Reactor Coolant Pump caused Reactor Trip from 100%. 50G (Overcurrent relays) were picked up on the Safety Breaker and the 6900V supply breaker. Automatic actuation of the motor driven (and turbine driven) Auxiliary Feed Pumps occurred as expected. All control rods fully inserted on the reactor trip. The steam generators water level is being maintained by the AFW system. Decay heat is being removed to the main condenser via the turbine dump valves. RCS PORV lifted and reseated. The electrical plant is in a normal shutdown lineup and the EDGs are available. There is no affect on Unit 2 due to this event. The licensee notified the NRC Resident Inspector. | Steam Generator Main Condenser Control Rod | |
| ENS 42211 | 17 December 2005 08:11:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip from Full Power Due to a Failed Feed Flow Channel | 'A' Steam Generator hi/hi level caused a turbine trip which in turn caused a Unit 1 reactor trip. The hi/hi steam generator level also caused both feedwater pumps to trip which caused the auxiliary feedwater pumps to auto-start. The hi/hi level also caused a feedwater isolation. Appropriate emergency procedures have been implemented and the plant is currently stable. All control rods fully inserted on the trip. Offsite power is available and powering safety related buses. The emergency diesel generators are available if required. There are no known primary to secondary leaks in the steam generators. The licensee will notify the NRC Resident Inspector. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 41578 | 8 April 2005 19:43:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Main Steamline Isolation Valves Closed on Valid Signal | The following information was provided by facsimile (licensee quotes in text): Valid actuation of main steam isolation valves that was not a part of pre-planned testing. Actuation was caused by going above pressurizer high pressure block (1955 psig) with steamline pressure less than 775 psig. Main steam isolation signal was reset. Main steam isolation valves reopened. No plant transient." The NRC Resident Inspector will be notified. | Main Steam Isolation Valve Main Steam | |
| ENS 41457 | 2 March 2005 21:29:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Motor Driven Afw Pumps During Outage | The following information was provided by the licensee via facsimile (licensee text in quotes): At 1629 (EST) on March 2, 2005, McGuire Nuclear Station Unit 2 experienced an automatic actuation of the 2A and 2B Motor Driven Auxiliary Feedwater (MDCA) Pumps. At the time of these actuations, Unit 2 was in MODE 3 in a refueling outage, the 2A Main Feedwater (CF) Pump was in service and, as per procedure, the 2B CF Pump was in a tripped condition and out of service. While performing system evolutions, Unit 2 CF system pressure unexpectedly dropped to a level that caused initiation of a low CF pump suction pressure signal which tripped the 2A CF Pump. Since the 2B CF Pump had previously been placed in a tripped condition, the logic for automatic start of the 2A and 2B MDCA Pumps was satisfied and these pumps started as designed. Since the 2A and 2B MDCA Pumps started in response to actual plant conditions or parameters satisfying the requirements for initiation of the Auxiliary Feedwater (CA) System safety function, this represented a valid actuation of the CA system reportable under the requirements of 10 CFR 50.72(b)(3)(iv)(A) . Note that the 2A and 2B MDCA Pumps started and operated normally and Unit 2 remained in a stable condition during this event. Unit 2 CF system parameters and alignments have been returned to normal for the current plant MODE and conditions. The 2A and 2B CA Pumps are now off. The licensee notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater |