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On March 9, 2016, at approximately 1034 hours, with Unit 1 at 100 percent power in Mode 1 and Unit 2 at 0 percent power in Mode 6 for a scheduled refueling outage, 2-QS-MOV-202B failed to stroke open when functionally tested due to excessive unseating thrust. An extent of condition review and engineering evaluation determined that 2-QS-MOV-202A maintained its function. An extent of condition review and engineering evaluation of the Unit 1 valves determined that 1-QS-MOV-102B maintained its function and 1-QS-MOV- 102A did not. Design Changes (DC) (DC NA-16-00023 for Unit 1 and DC NA-16-00021 for Unit 2) were implemented to provide additional valve operating margin prior to returning the valves to service. Based on as- found testing and engineering evaluation, Sodium Hydroxide (NaOH) injection for both units would have been supplied in the event of a Design Basis Accident (DBA) and the safety function of the Chemical Addition System was maintained. The health and safety of the public were not affected by this event.

.NRC FORM 366 (11-2015) APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 Reported lessons learned are incorporated into the licensing process and fed back to industry. Send.

comments' regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officet Office of Information and Regulatory Affairs, used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

On February 26, 2015, at 1511 hours with Unit 1 operating at 96 percent power in an end of cycle coastdown, an automatic reactor trip occurred. The initiating signal was a low-low level on "B" steam generator caused by the closure of the "B" main feedwater regulating valve (MFRV). Closure of the valve was due to a loss of power on the final driver card for "B" MFRV. At 1639 hours, a 4-hour Non-Emergency Report was made to the NRC in accordance with 10 CFR 50.72(b)(2)(iv)(B) for "an event causing actuation of the Reactor Protection System when the reactor is critical" and an 8-hour report was also made in accordance with 10 CFR 50.72(b)(3)(iv)(A) for "an event causing actuation of the Auxiliary Feedwater System (AFW).

Approximately 30 minutes after the reactor trip with AFW flow throttled, the Turbine Driven (TD) AFW pump, 1-FW-P-2, discharge relief valve (RV) lifted and discharged approximately 200 gallons per minute (gpm) to the ground. The RV opened as a result of the inability of the governor valve for the pump to travel an additional 3/16" in the closed direction. An engineering evaluation determined that the loss of emergency condensate storage tank inventory was a condition reportable per 10 CFR 50.73(a)(2)(i)(B) for an "any operation or condition which was prohibited by Technical Specifications." The health and safety of the public were not affected by either event.

At 0859 on February 2, 2014, with Unit 2 operating at 100% power, the "A" Math Feedwater (MFW) Pump, 2-FW- P-1A, had a motor lead connection that grounded and caused the power supply breakers for 2-FW-P-1A to trip open. The standby MFW pump, 2-FW-P-1C, auto started, as designed. The Reactor Operator (RO) believed the standby pump did not auto start, since one of the Control Room indicating lights did not illuminate as expected.

The RO believed that only one MFW pump was running with two required for operation greater than 70% power. Based on this indication and the perceived loss of MFW, the RO initiated a manual reactor trip as directed in 2-AP-31. Although the remaining control room team members were aware that the standby pump had auto- started, they were unable to intervene in time to prevent the manual trip. Following the manual reactor trip, the unit was stabilized in Mode 3 at normal reactor coolant temperature and pressure. This event was reportable per 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system. Following the reactor trip, the Auxiliary Feedwater pumps automatically started as designed and provided makeup flow to the steam generators. This event was reportable per 10CFR50.72(b)(3)(iv)(A) for actuation of an Engineered Safety Feature (ESF) system. The health and safety of the public were not affected by this event.

On September 15, 2014, with Unit 2 defueled, debris that had the potential to be fuel fragments was located on the core plate directly below the B11 core location, where fuel assembly 4Z9 resided during Cycle 23.

Video inspection of fuel assembly 4Z9 identified that the top springs of two fuel pins were dislodged. Due to the fact that the fuel damage exceeded expected conditions, at 1454 on September 15, 2014, this event was reported as an eight hour report as per 10 CFR 50.72(b)(3)(ii)(A), any event or condition that results in the condition of the nuclear plant, including its principle safety barriers, being seriously degraded. Detailed video inspections estimated that 15 fuel pellets were dislodged from fuel assembly 4Z9. During efforts to identify and recover the fuel pellets, 7 fuel pellets worth of material were not found and have already or are expected to granulate into fine particles that will dissolve in low flow areas of the primary plant systems, or be removed by normal purification processes. Since the specific location of the 7 fuel pellets is undesignated and because those pellets contain licensed material in a quantity greater than 10 times the quantity specified in App. C of 10 CFR 20, a report was made at 1227 on September 30, 2014, pursuant to 10 CFR 74.11(a) and to 10 CFR 20.2201(a)(ii). The health and safety of the public were not affected by this event.

On October 11, 2013, at 1319 hours with Unit 1 operating at 48 percent power (Mode 1), an automatic turbine trip and subsequent reactor trip occurred due to a lockout relay actuation for the 1C Station Service Transformer (1-EP-SST-1C). The lockout occurred simultaneously with the start of the 1C Condensate Pump (1-CN-P-1C). The direct cause of the 1-EP-SST-1C lockout is that current transformer terminal block shorting screws were left installed inside the 1- EP-BKR-15C2 breaker cubicle. The root cause of the event was less than adequate written instructions for documenting the installation and removal of the terminal block shorting screws.

All safety system responded as expected. The Auxiliary Feedwater Pumps actuated as designed following the reactor trip and provided makeup flow to the Steam Generators. 1-EP- SST-1C was inspected and no signs of damage or abnormal conditions were observed. At 1507 hours, a 4 hour report was made to the NRC in accordance with 10CFR50.72(b)(2)(iv)(B) for a Reactor Protection System (RPS) actuation and a 8 hour report in accordance with 10CFR50.72(b)(3)(iv)(A) for a Auxiliary Feedwater system actuation. The event is reportable pursuant to 10CFR50.73(a)(2)(iv)(A) for a condition that resulted in the automatic actuation of the RPS and AFW Systems. The health and safety of the public were not affected by the event.

On February 3, 2011, at 0345 hours, with Unit 1 at 100 percent power, Mode 1, annunciator 1H-G4, Annunciator System DC Ground was received in the Main Control Room (MCR).

At 0348 hours annunciator 1B-D3, Boric Acid Tank 1B Hi-Lo Level CH I-II was received which, when acknowledged, locked in and annunciator 1H-G4 cleared. While investigating, an acrid smell was noticeable in the MCR. At 0353, upon entry into the annunciator system cabinet room, adjacent to the MCR, the door of the cabinet, 1-EI-CB-21, was opened and flames approximately 2 - 4 inches long were observed coming from an annunciator circuit card. Operations personnel entered fire contingency action procedure 0-FCA-0, Fire Protection - Operations Response. At 0354 hours, a two second discharge of a portable CO2 fire extinguisher put out the fire and a re-flash fire watch was established. At 0437 hours the fire contingency action procedure was exited. This event posed no significant safety implications since the fire was small, extinguished quickly, did not pose an actual threat to the safety of the nuclear power plant and it did not affect equipment required for safe operation of the plant. Therefore, the health and safety of the public were not affected by this event. This event is being reported voluntarily to share information and lessons learned.

On February 8, 2008, Unit 2 was in Mode 3, zero percent power, preparing for restart following a planned maintenance outage. At 1759 hours, a rod control urgent failure alarm was received while withdrawing the "A" Shutdown Bank control rods. Rod withdrawal was stopped and it was noted that the Group Step Counters deviated by three steps.A As a result, at 1812 hours a manual reactor trip was initiated by opening the reactor trip breakers in accordance with Technical Requirement 3.1.3. This is a valid actuation of the Reactor Protection System. Four suspect cards were replaced in Power Cabinet 2AC.

Following card replacement the reactor trip breakers were closed.A Shutdown Banks "A" and "B" were fully withdrawn with no concerns. On February 8, 2008, at 1959 hours, a non- emergency 8-hour report was made in accordance with 10 CFR 50.72(b)(3)(iv)(A) due to actuation of the Reactor Protection System. This event is reportable pursuant to 10 CFR 50.73A (a)(2)(iv)(A) for aA condition thatA resulted inAA manualA actuationA of theA Reactor Protection System. This event posed no significant safety implications because the reactor was subcritical when the reactor trip breakers were opened. Compliance with all Technical Requirements was maintained. Therefore, the health and safety of the public were not affected by this event.

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On October 3, 2005, with Unit 2 in Mode 5 for a scheduled refueling outage it was determined that all required actions for Low Temperature Overpressure Protection (LTOP) System controls were not implemented. A second independent means to prevent more than one low head safety injection (LHSI) pump and charging pump from being capable of injecting into the Reactor Coolant System (RCS) was not being performed. The procedures controlling the LHSI and charging pumps operations only required the pumps to be placed in pull to lock.

The Technical Specifications (TS) Bases states two independent means are required to prevent a pump start such that a single failure or single action will not result in an injection to the RCS. This condition was applicable to Units 1 and 2. This event is reportable pursuant to 10 CFR 50.73 (a)(2)(i)(A) for a condition prohibited by the TS. This event posed no significant safety implications since the LTOP System design basis pressure and temperature limit curve were never violated. Therefore, the health and safety of the public were not affected by this event.