05000251/LER-2015-001

LER-2015-001, Automatic Auxiliary Feedwater System Actuation during a Planned Reactor Trip

Turkey Point Unit 4
Event date:	11-30-2014
Report date:	1-29-2015
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation
Initial Reporting
ENS 50645	10 CFR 50.72(b)(3)(iv)(A), System Actuation
2512015001R00 - NRC Website

DESCRIPTION OF THE EVENT

On November 30, 2014, Turkey Point Unit 4 reactor [AC, RCT] was in Mode 1, with reactor power reduced to 23% to facilitate the discovery and repair of an unidentified steam leak on the Unit 4 High Pressure (HP) Turbine. Unable to determine the exact location of the HP turbine steam leak, at approximately 1354 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.15197e-4 months <br />, the Unit 4 reactor was manually tripped in accordance with normal operating procedure guidance, as a pre-planned evolution. The planned Unit 4 reactor shutdown was conducted in accordance with Operating Procedure 4-G0P-103, Power Operation to Hot Standby.

Following the reactor trip, as expected, Operators entered Emergency Operating Procedure 4-EOP-E-0, "Reactor Trip or Safety Injection," and transitioned to 4-E0P-ES-0.1, "Reactor Trip Response," to stabilize and control the plant following a reactor trip without a safety injection.

As part of the reactor trip response, 4-E0P-ES-0.1, Step 2b, Operators were verifying that the Main Feedwater Control Valves [SJ,FCV] had closed, when they observed dual indication on the 4A main feedwater flow control valve, 4-FCV-478. This unexpected indication forced the operating crew to manually close the Main Feedwater Control Valves, by performing the step in the "Response Not Obtained" (RNO) column. This unexpected manual action, delayed the expeditious performance of subsequent steps in the procedure to establish feedwater flow using the Main Feedwater Bypass Valves, as directed in 4-E0P-ES-0.1 Step 2f.

At approximately 1358 hours0.0157 days <br />0.377 hours <br />0.00225 weeks <br />5.16719e-4 months <br />, following the reactor trip, Unit 4 was in Mode 3 when the level in the 4C Steam Generator (SG) [SB,SG] level decreased to the low-low level setpoint, 16% Narrow Range (NR), causing an Auxiliary Feedwater (AFW) System [BA] actuation.

At approximately 1407, the Main Steam Isolation Valves [SB,ISV] were procedurally closed in response to the cool down. AFW was restored to standby alignment at approximately 1454 hours0.0168 days <br />0.404 hours <br />0.0024 weeks <br />5.53247e-4 months <br />. The 4A Main Feedwater Pump was secured at 1526 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.80643e-4 months <br />, with the 4A Standby Feedwater Pump supplying feed to the SGs and decay heat removed via the atmospheric dump valves.

The NRC Operations Center was notified by Event Notification 50645 in accordance with 10 CFR 50.72(b)(3)(iv)(A) for valid actuation of the AFW system.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as "...any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section." The AFW System was automatically actuated during the event and is included in the systems listed in paragraph 10 CFR 50.73(a)(2)(iv)(B).

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION Net / LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

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 intemet 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.

CAUSE OF THE EVENT

The causal analysis determined the following:

• The appropriate operating margin to prevent AFW actuation was not established prior to the reactor trip for the planned shutdown.

• The just-in-time training did not prepare crews to reduce the probability of having an unnecessary AFW actuation on a planned reactor trip.

ANALYSIS OF THE EVENT

On November 28, 2014, an unidentified steam leak was discovered on the south end of the Unit 4 HP turbine. Due to an excessive amount of steam, personnel could not gain access to identify the source of the steam leak. A load reduction was necessary to allow personnel to enter the area but visibility and accessibility was limited. On November 30, the Unit 4 was manually tripped from 23% power.

Following the trip, the 4C SG level decreased to 16% NR, the low-low level setpoint, and AFW actuated.

Providing feedwater through the main feedwater bypass valves to the generators would have been sufficient to maintain reactor coolant temperature stable after the trip, since Unit 4 was at the beginning of cycle (BOC) plant conditions with low decay heat load.

On a planned shutdown, reducing steam demand prior to tripping the reactor and the turbine reduces the SG level shrink, and therefore reduces the probability of an AFW actuation following the reactor trip. The operating procedure, 4-GOP-103, permitted Operators to trip the reactor manually when reactor power decreases to approximately 15 to 25% power. However, prior to that step, the procedure includes a note to caution the Operators that a " Manual trip below 20% power reduces the probability of unnecessary Auxiliary Feedwater Actuation and enables more effective control of steam generator levels.

For this event, the reactor was tripped from 23% power, in order to minimize the risk of having a secondary transient due to the unidentified HP turbine steam leak, which increased the probability of an AFW actuation. This operating margin reduction, along with the delay in the expeditious performance of the procedural steps in 4-E0P-ES-0.1 (due to the unexpected dual indication observed at FCV-4-478 controls), resulted in not establishing feedwater flow to the generators using the Feedwater Bypass Valves promptly, reaching the SG low-low level setpoint setting of 16% NR, and thus actuating AFW.

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION f t ',1/47kire, LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

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 intemet 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.

ANALYSIS OF SAFETY SIGNIFICANCE

AFW initiated on a low-low SG level signal and added cooler water to the generators. As expected, Operators closed the Main Steam Isolation Valves to control the cooldown. AFW was later secured.

During this event, Operator actions were successful in controlling the cooldown, stabilizing the plant and maintaining the reactor in a safe condition. As such, the safety significance of this event is very low.

CORRECTIVE ACTIONS

Corrective actions are in accordance with condition report AR 2009853 and include:

1. Change the applicable operating procedures to establish available margin to avoid unnecessary AFW actuation during a planned reactor trip.

2. Develop simulator scenarios that more closely model the plant response during a planned shutdown and train Operators to reduce the probability of an unnecessary AFW actuation during a planned reactor trip.

ADDITIONAL INFORMATION

ENS Codes are shown in the format [IEEE system identifier, component function identifier, second component function identifier (if appropriate)].

FAILED COMPONENTS IDENTIFIED: None PREVIOUS SIMILAR EVENTS: None