05000391/LER-2023-003, Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control

Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control

ML23276A571
Person / Time
Site:	Watts Bar
Issue date:	10/03/2023
From:	Anthony Williams Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
WBL-23-045 LER 2023-003-00
Download: ML23276A571 (1)

LER-2023-003, Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation
3912023003R00 - NRC Website

text

TENNESSEE VALLEY AUTHORITY Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381 WBL-23-045 October 3, 2023 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-391 10 CFR 50.73 Subject: Licensee Event Report 391/2023-003-00, Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control Pursuant to the reporting requirements of 10 CFR 50.73, attached is the subject Licensee Event Report concerning the Automatic Reactor Trip and Reactor Protection System Actuation for Watts Bar Nuclear Plant, Unit 2, which occurred on August 4, 2023. A supplemental LER will be submitted after detailed causal evaluations are complete.

There are no new regulatory commitments contained in this letter. Please direct any questions concerning this matter to Jonathan Johnson, WBN Licensing Manager, at jtjohnson0@tva.gov.

Respectfu y.,--

Anthony, L--

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Site Vice President Watts Bar Nuclear Plant U.S. Nuclear Regulatory Commission WBL-23-045 Page 2 October 3, 2023

Enclosure: LER 391/2023-003-00, Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control

cc (w/Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Region II

ENCLOSURE Tennessee Valley Authority Watts Bar Nuclear Plant Unit 2

LER 391/2023-003-00, Automatic Reactor Trip Due to Loss of Main Feedwater Regulating Valve Control

Abstract

On August 4, 2023 at approximately 1746 Eastern Daylight-Saving Time (EDT), while both units were at 100 percent rated thermal power, Watts Bar Nuclear Plant (WBN) Unit 2 experienced an automatic reactor trip in response to controls failure of the main feedwater regulating valve to steam generator 2. All control and shutdown bank rods inserted properly in response to the reactor trip. All safety systems, including the Auxiliary Feedwater (AFW) System, performed as designed and there were no complications with the trip.

The most likely cause for the reactor trip was a failure of the associated Distributed Control System (DCS) field bus module (FBM) which was replaced. Detailed causal evaluations are ongoing and a supplemental LER will be submitted when additional information becomes available.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A), as an automatic actuation of the Reactor Protection System (RPS) and Auxiliary Feedwater System (AFW).

I. Plant Operating Conditions before the Event

Both Watts Bar Units 1 and 2 were at 100 percent Rated Thermal Power (RTP).

A. Event Summary

On August 4, 2023 at approximately 1746 Eastern Daylight-Saving Time (EDT), while both units were at 100 percent rated thermal power, Watts Bar Nuclear Plant (WBN) Unit 2 experienced an automatic reactor trip in response to a loss of feedwater regulating valve

[EIIS:FCV] control and a subsequent low steam generator (SG)[EIIS:SG] No. 2 level signal. All control and shutdown bank rods inserted properly in response to the reactor trip. All safety systems, including the Auxiliary Feedwater (A FW)[EIIS:BA] System, performed as designed and there were no complications with the trip. Based on investigation and troubleshooting following the event, the most probable cause was a failure of a Distributed Controls System (DCS)[EIIS:JB] field bus module (FBM) that provides data input to feedwater valve controllers

[EIIS:FC]. The FBM in question and its pair were subsequently replaced and sent off for failure analysis. A supplemental LER will be submitted when the causal evaluations and analyses are complete.

Notification to the NRC of the reactor trip was made by Operations at 2051 EDT (Event Notification 56660). This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) due to actuation of the Reactor Protection System (RPS) and Auxiliary Feedwater (AFW) System.

B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event

There were no safety related inoperable structures, components, or systems that contributed to this event.

C. Dates and approximate times of occurrences

Dates and Approximate Times Occurrence Received DCS Trouble Alarm with SG flow low alarm in.

SG No.2 bypass flow was noted to be high on scale and SG 8/04/2023 at 1742 EDT No. 2 level was noted to be approximately 30 percent lowering. Using manual controls, SG No. 2 level was restored to approximately 60 percent. Entered 2-AOI-16, Loss of Normal Feedwater.

Unit 2 Reactor Tripped on Low Low SG Level. Unit 2 8/04/2023 at 1746 EDT entered Mode 3. Feedwater flow unable to be maintained with SG No. 2 Feedwater Regulating Valve (FRV) in manual. Exited 2-AOI-16 8/04/2023 at 1750 EDT Entered ES-01, Reactor Trip Response

8/04/2023 at 1803 EDT Entered 2-GO-5, Unit Shutdown from 30 percent Reactor Power to Hot Standby

D. Manufacturer and model number of each component that failed during the event

Foxboro, Field bus module, FBM218 (P/N: P0922VW)

E. Other systems or secondary functions affected

None.

F. Method of discovery of each component or system failure or procedural error

Initial event troubleshooting identified a faulty field bus module (FBM) within the Distributed Control System (DCS). Additional information will be provided in a supplemental LER.

G. The failure mode, mechanism, and effect of each failed component

The failed field bus module caused a loss of main feedwater regulating valve control. The exact failure mechanism of the module is unknow n at this time. Additional information will be provided in a supplemental LER.

H. Operator actions

Operations personnel promptly stabilized the plant following the reactor trip.

I. Automatically and manually initiated safety system responses

The automatic reactor trip was followed by an automatic subsequent turbine trip. Safety systems responded as expected.

A. Cause of each component or system failure or personnel error

The exact cause of the FBM component failure is unknown, pending further analysis.

Additional information will be provided with a supplemental LER.

B. Cause(s) and circumstances for each human performance related root cause

No human performance issues were attributed to this event.

IV. Analysis of the event

While operating at approximately 100 percent power during steady state operation, the Watts Bar Unit 2 reactor experienced an automatic reactor trip at approximately 1746 EDT on August 4, 2023, in response to low SG No. 2 level. This automatic Unit 2 reactor trip was comparable to the Updated Final Safety Analysis Report (UFSAR) transients/accidents. The sequence of events associated with the trip were bounded by the FSAR Safety Analysis assumptions. The plant response post -trip was uncomplicated and the plant responded as designed. Operations entered 2-AOI-16, Loss of Normal Feedwater; 2-E-0, Reactor Trip or Safety Injection and subsequently transitioned to 2 -ES-0.1, Reactor Trip Response and 2 -GO-5, Unit Shutdown from 30 percent Reactor Power to Hot Standby.

V. Assessment of Safety Consequences

In review of the associated plant system performance data from the key plant parameters and plant response to the transient it was concluded that all safety systems performed as designed. There were no conditions identified where actuation signals were not appropriately generated based on the plant transient. The parameter response for this automatic reactor trip was bounded by the FSAR safety analysis assumptions. Probability Risk Assessments (PRA) performed for this Unit 2 trip event resulted in very small impacts to core damage frequency (CDF) and large early release frequency (LERF) values.

A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event

Not applicable.

B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident

Not applicable.

C. For failure that rendered a train of a safety system inoperable, estimate of the elapsed time from discovery of the failure until the train was returned to service

Not applicable.

VI. Corrective Actions

This event was entered into the TVA Corrective Action Program and is being tracked under Condition Report 1872637.

A. Immediate Corrective Actions

Operations personnel promptly stabilized the plant in Mode 3. Troubleshooting and field inspections identified the faulty field bus module which was subsequently replaced.

B. Corrective Actions to Prevent Recurrence or to reduce the probability of similar events occurring in the future

A detailed causal analysis of the failed component is being performed. Additional information will be provided with a supplemental LER.

VII. Previous Similar Events at the Same Site

LER 391-2023-002, Automatic Reactor Trip Due to Main Generator Protection Relay Actuation documents a Unit 2 reactor trip that occurred on June 27, 2023. This event was caused by a loose terminal connection for current input into a generator protection relay.

LER 390-2020-001, Manual Reactor Trip Due to Lowering Steam Generator Level Caused by a Hand Station Failure documents a Unit 1 manual reactor trip that occurred on February 2, 2019. This event was caused by a stuck pushbutton on the main feedwater regulating valve controller/hand station which led to a loss of control for steam generator No. 3 level.

The LERs listed above do not share a similar cause to the reactor trip event described in this LER.

VIII. Additional Information

None.

IX. Commitments

None.