05000423/LER-2016-005

LER-2016-005, Technical Specification Required Shutdown and Manual Reactor Trip Due to Steam Generator Level Oscillation

Millstone Power Station Unit 3
Event date:	06-12-2016
Report date:	08-09-2016
Reporting criterion:	10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation
4232016005R00 - NRC Website

LER 16-005-00 for Millstone Power Station Unit 3 RE: Technical Specification Required Shutdown and Manual Reactor Trip Due to Steam Generator Level Oscillation

ML16230A004
Person / Time
Site:	Millstone
Issue date:	08/09/2016
From:	Daugherty J R Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
16-321 LER 16-005-00
Download: ML16230A004 (6)

1. EVENT DESCRIPTION:

On June 12, 2016 at approximately 19:55, with Millstone Power Station Unit 3 (MPS3) operating in MODE 1 at 100% power, operators identified the third stage on the "A" Reactor Coolant Pump (RCP) seal had failed, which resulted in an unidentified Reactor Coolant System (RCS) leak greater than the Technical Specification (TS) limit. The leak was going to the containment sump at a leakage rate of approximately 2 gallons per minute (gpm) which exceeded the TS 3.4.6.2.b limit of 1 gpm UNIDENTIFIED LEAKAGE. Operators entered TS 3.4.6.2 ACTION Statement b, "With any RCS operational LEAKAGE not within limits, other than PRESSURE BOUNDARY LEAKAGE, LEAKAGE from Reactor Coolant System Pressure Isolation Valves or primary to secondary LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />", and commenced a downpower at 1% per minute in accordance with station procedures. During the downpower, steam generator levels were not adequately maintained. As a result, the Engineered Safety Features Actuation System generated a Turbine Trip and Feed Water Isolation on Steam Generator Water Level — High-High being exceeded on the 'B' steam generator. In response to the turbine trip and feedwater isolation, operators manually tripped the reactor at 23:37 (MODE 1, at approximately 20 )`/0 power). The auxiliary feedwater system started as designed and maintained steam generator levels. Safety systems functioned as expected. There were no radiological challenges as a result of the event. The plant completed the Shutdown required by Technical Specifications and entered COLD SHUTDOWN on June 14, 2016 at 01:29.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B).

Additionally, the plant shutdown is being reported in accordance with 10 CFR 50.73 (a)(2)(i)(A) as the completion of any nuclear plant shutdown required by the plant's Technical Specifications.

2. CAUSE:

The cause of the UNIDENTIFIED LEAKAGE exceeding plant technical specification requirements was a failed third stage on the "A" RCP seal. Failure of the "A" RCP seal stage was caused by gradual pitting and degradation of the seal stationary faces.

The cause of the Engineered Safety Features Actuation System generated turbine trip and feedwater isolation was inadequate control of feedwater flow by the feedwater station operator which resulted in all four steam generator levels increasing and the "B" SG reaching the 80% Hi- Hi trip setpoint.

3. ASSESSMENT OF SAFETY CONSEQUENCES:

The failure of the third stage of the 'A' RCP seal followed by an excess feedwater flow transient and manual reactor trip during the rapid downpower, was bounded by the analysis presented in Final Safety Analysis Report Chapter 15.1.2, "Feedwater System Malfunctions that Result in an Increase in Feedwater Flow". Therefore, the event had a very low safety significance.

The leak resulted from failure of only the upper third stage of the three stage Flowserve RCP seal design. Pressure breakdown and flow restriction from the first two seal stages continued. The resultant leak rate from the "A" RCP seal, approximately 2 gpm, was within the capability of the charging system to maintain pressurizer level. The leakage from the seal was collected in the containment sump with no releases to the environment.

4. CORRECTIVE ACTION:

The 'A' RCP seal was replaced. Additional corrective actions will be taken in accordance with the corrective action program.

Operator performance issues for the impacted positions are being addressed in the Corrective Action Program.

5. PREVIOUS OCCURRENCES:

There are no previous occurrences with the same underlying reason or consequences.

6. ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES:

• Reactor Coolant System — AB

• Containment- NH

• Feedwater System — JB

• Isolation Valve — ISV

• Pump — P

• Solid State Protection System — JG

• Steam Generator- SG