05000348/LER-2016-006

LER-2016-006, Manual Reactor Trip due to Loss of Speed Control on 1A Steam Generator Feed Pump

Joseph M. Farley Nuclear Plant
Event date:	11-08-2016
Report date:	12-19-2016
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation
LER closed by
IR 05000348/2017001 (1 May 2017)
3482016006R00 - NRC Website

LER 16-006-00 for Joseph M. Farley Nuclear Plant, Unit 1 Regarding Manual Reactor Trip due to Loss of Speed Control on 1A Steam Generator Feed Pump

ML16354B547
Person / Time
Site:	Farley
Issue date:	12/19/2016
From:	Gayheart C A Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-16-2480 LER 16-006-00
Download: ML16354B547 (6)

A. PLANT AND SYSTEM IDENTIFICATION

Westinghouse - Pressurized Water Reactor

B. DESCRIPTION OF EVENT

At 1326, on November 8, 2016, reactor power was at 34 percent in Mode 1 and power was being reduced to remove the main generator from service. The recirculating flow control valve [FCV] for the 1A Steam Generator Feed Pump (SGFP) [P] was opened in accordance with unit shutdown procedure. All steam generator (SG) [SG] levels began to trend down, as expected and as described in the unit operating procedure. The unit operator attempted to raise speed for the 1A SGFP using the SGFP master speed controller [SC]. The SGFP speed rose to approximately 3700 rpm but would not rise further. The demand on the SGFP master speed controller was greater than 65 percent, which corresponded to a speed greater than 3700 rpm.

Due to the inadequate response from the SGFP master speed controller, the Shift Supervisor directed raising speed on the 1A SGFP using the slave controller. Demand on the slave controller was raised to approximately 80 percent, which corresponded to a speed higher than 3700 rpm. The 1A SGFP speed was observed to not have changed and remained at 3700 rpm.

The Shift Supervisor then established trip criteria. Due to the inability to raise 1A SGFP speed and feedwater flow to the SGs, all SG levels continued to lower. At 32 percent reactor power, the Shift Supervisor directed a manual reactor trip to prevent reaching the low SG level automatic reactor trip setpoint. At 1331, a manual reactor trip was performed due to lowering SG levels and an inability to raise speed and feedwater flow because the 1A SG FP speed control circuitry failed. The motor driven auxiliary feed pumps also started automatically, as expected, with the manual reactor trip. The main steam isolation valves were closed to limit the unit cool down because the isolation valves [ISV] on the 2A/2B moisture separator reheater [MSR] second stage had excessive Ieakby. Decay heat removal was accomplished with the atmospheric relief valves, and the unit was maintained in mode 3 to continue with the previously scheduled main generator repairs.

C. UNIT STATUS AT TIME OF EVENT

Unit 1, Mode 1, 32 percent power with a unit shutdown in progress

D. CAUSE OF EVENT

The speed reference adjust and speed controller (C2) card was found to have an out of tolerance output voltage due to an A2 operational amplifier on the card that failed because of infant mortality.

E. REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT

This event is reportable as required by 10 CFR 50. 73(a)(2)(iv)(A) due to a manual actuation of the reactor protection system and automatic actuation of the auxiliary feedwater system. The reactor was shut down at 1331 and mode 3 was entered. There was no loss of safety function internet e-mail to Infocoffects.ResourceOnrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEDB-10202, (3.150-0104), Office of Management and Buget.

Washington, DC 20503. If a means used to Vic' Al in; forrroabon ca'Iectun does not &splay a cwrentty valid OMB control number, the NRC may root conduct a sponsor, and a person is not reared to respond to, the information collection.

Joseph M. Farley Nuclear Plant, Unit 1 05000- 348 ,A1 .. , Fr and no radioactive release associated with this event. All required safety systems were available and responded as expected. There were no actual consequences detrimental to the health and safety of the public and the event is considered to be of very low safety significance.

F. CORRECTIVE ACTION

A new C2 card was installed, and the 1A SGFP high pressure and low pressure governor valves were stroked satisfactorily as the functional test.

G. ADDITIONAL INFORMATION

1) Previous Similar Events: none 2) Commitment Information: No commitments are made in this correspondence.