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 Report dateSiteEvent description
05000370/LER-2017-00126 June 2017McGuire

On February 23, 2017, at 19:22 hours, with Unit 1 and Unit 2 operating at approximately 100 percent power, operators commenced a Unit 2 shutdown upon discovery of pressure boundary leakage on Unit 2 Safety Injection (NI) pipe upstream of the connection to "D" Reactor Coolant System (NC) Cold Leg. During a containment walk down inspection in Mode 3 on the next day, a pinhole pressure boundary leak was observed in the body of 2NC-30, Pressurizer Spray Bypass Valve.

The cause of the NI pipe leak is thermal fatigue damage caused by NC cross-loop flows. The cause of the 2NC-30 valve leak is a casting flaw attributed to a combination of defects during the manufacturing process that resulted in a through wall pinhole leak in the valve body. The NI pipe with the flaw and the valve with the pinhole leak could have structurally performed their design functions. Therefore, the health and safety of the public were not affected by these events.

Valve 2NC-30, the NI pipe, and leaking B-Loop NI check valves were replaced. Thermal cycling monitoring and mitigation devices were installed on Unit 2 and will be installed on Unit 1 during the next refueling outage.

05000369/LER-2016-00122 July 2016McGuire
Mcguire

On March 22, 2016, while Unit 1 was in end of cycle (EOC) refueling outage 1E0C24 (Mode 5), a manual ultrasonic (UT) examination of the Chemical and Volume Control System (NV), Charging Line to the 1A Reactor Coolant System (NC) cold leg confirmed a previously identified circumferential indication associated with weld 1NC1F-1374. The current examination results had shown that the indication had changed since the previous examination during 1E0C23 and concluded that the indication no longer met American Society of Mechanical Engineers (ASME) Section XI Code requirements. This condition is reportable under 10CFR50.73(a)(2)(ii)(A) as a degraded condition.

A specific cause for the condition could not be determined. A metallurgical examination concluded that the cause of the UT result could have been influenced by pre-existing welds associated with a legacy modification.

The affected NV piping on Unit 1 was replaced during refueling outage 1E0C24.

APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 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 internet e-mail to infocollects Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

05000370/LER-2015-0017 December 2015McGuire

On October 7, 2015, an actuation of the Auxiliary Feedwater (CA) system occurred while Unit 2 was in Mode 4 and operators were restoring from testing of the 2A Solid State Protection System (SSPS) Safety Injection (SI) trip functions for the Main Turbine and the Main Feedwater (CF) pump turbines. The CA actuation caused the 2A CA Train flow control valves to fully open and the associated steam generator (SG) sampling and blowdown valves to close. The actuation occurred as designed, and there was no adverse impact to plant operation.

At the time of the CA actuation, the CF pumps were shut down and SG levels were being maintained by the 2A and 2B Motor Driven CA Pumps (MDCAP)s. During restoration from SSPS testing, a conditional step in the procedure did not clearly require the reset of at least one CF pump. Leaving both CF pumps in the tripped state provided the logic for the CA actuation signal. The actuation occurred when the 2A Train CA auto start defeat switch was placed in "reset" as directed by the test procedure. To prevent the CA actuation, the test procedure should have ensured that at least one CF pump was in "reset" before the 2A CA Train auto start defeat switch was placed in "reset".

Procedures PT/1/A/4200/026A and PT/2/A/4200/026A will be revised to clearly define the CF pump as found and restoration configuration based on actual plant conditions.

APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 Estimated burden per response to Comply with this mandatory collection request: 80 hours.

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 internet e-mail to infocollects Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

05000369/LER-2013-0036 March 2014McGuire

On November 14, 2013, at 13:13, Unit 1 was manually tripped from 100% power due to 10 dropped control rods associated with Rod Control Power Cabinet 1AC. The remaining control rods fully inserted into the core following the manual reactor trip. Unit 1 was stabilized in Mode 3 at normal operating temperature and pressure. The motor-driven Auxiliary Feedwater Pumps 1A and 1 B were manually started for steam generator level control. This event did not impact public health and safety.

The cause of the event was an inadequate modification, resulting in an over-voltage protection (OVP) setpoint too close to the normal output of both the primary and backup -24 VDC rod control power supplies. The design called for the OVP function to be implemented with an installed jumper configuration that set the OVP setpoint too close to the normal output voltage.

Actions were taken to replace other Unit 1 power supplies with available power supplies that will not shut down under similar conditions. Due to limited spares, immediate replacement was limited to all applicable primary supplies and two backup supplies on Unit 1. The remaining backup power supplies installed in Unit 1 and the susceptible Unit 2 power supplies will be replaced as part of the planned corrective actions.

366A U.S. NUCLEAR REGULATORY COMMISSION