05000458/LER-2023-005, Manual Reactor Scram Due to Lowering Feedwater Temperature Following an Automatic Isolation of Feedwater Heater String

Manual Reactor Scram Due to Lowering Feedwater Temperature Following an Automatic Isolation of Feedwater Heater String

ML24016A189
Person / Time
Site:	River Bend
Issue date:	01/16/2024
From:	Crawford R Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RBG-48270 LER 2023-005-00
Download: ML24016A189 (1)

LER-2023-005, Manual Reactor Scram Due to Lowering Feedwater Temperature Following an Automatic Isolation of Feedwater Heater String

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation
4582023005R00 - NRC Website

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} entergy RBG-48270 January 16, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Randy Crawford Manager Regulatory Assurance 225-381-4177 10 CFR 50.73 Subject: Licensee Event Report 50-458 I 2023-005-00, Manual Reactor Scram Due to Lowering Feedwater Temperature Following an Automatic Isolation of Feedwater Heater String River Bend Station -Unit 1 NRC Docket Nos. 50-458 Renewed Facility Operating License No. NPF-47 In accordance with 10 CFR 50. 73, enclosed is the subject Licensee Event Report.

This document contains no commitments.

Should you have any questions, please contact Mr. Randy Crawford, Regulatory Assurance Manager, at 225-381-4177.

Respectfully, Randy Crawford RC/db Enclosure:

cc: Licensee Event Report 50-458 / 2023-005-00, Manual Reactor Scram Due to Lowering Feedwater Temperature Following an Automatic Isolation of Feedwater Heater String NRC Region IV Regional Administrator -Region IV NRC Senior Resident Inspector -River Bend Station Entergy Operations, Inc., 5485 U.S. Highway 61 N. St Francisville, LA 70775 -µ r

Enclosure

RBG-48270

Licensee Event Report 50-458 / 2023-005-00, Manual Reactor Scram Due to Lowering Feedwater Temperature Following an Automatic Isolation of Feedwater Heater String 1

Abstract

On November 17, 2023, at 22:15 CST, with River Bend Station, Unit 1, operating in Mode 1 at 30% power, an isolation of Low-Pressure Feedwater Heater String A occurred. Operators entered the applicable Abnormal Operating Procedure (AOP) and reduced power to mitigate the lowering feedwater temperature. Reactor power was lowered to 24% and a manual reactor scram was inserted at 23:55 CST. All control rods fully inserted and there were no complications. All systems responded as designed.

The cause of the Heater String isolation is still under investigation. A supplement to this report will be issued once the investigation is complete.

This report is made pursuant to 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of the Reactor Protection System.

EVENT DESCRIPTION

On November 17, 2023, at 22:15 CST, River Bend Station (RBS), Unit 1, was operating at 30% power while performing plant startup activities when an isolation of Low-Pressure Feedwater Heater String A [SJ] occurred. Main Control Room Operators entered Abnormal Operating Procedure - Loss of Feedwater Heating (AOP-0007) for transients involving lowering feedwater temperature. AOP-0007, Attachment 1 (Feedwater Temperature vs Core Thermal Power Graph),

provides guidance for power operations with reduced feedwater temperature from 100% reactor power to 25% reactor power. In accordance with AOP-0007, Attachment 1, operators began to manually insert control rods using the Control Rod Drive System [AA] to exit the restricted region for power operations at the lower feedwater temperature. Control rod insertion lowered power to 24%, which placed the unit below the graph start point of AOP-0007, Attachment 1. At 23:55 CST, while RBS was operating at 24% power, a manual reactor scram was inserted due to operation below the graph start point of AOP-0007, Attachment 1, caused by continued lowering of feedwater temperature. All control rods fully inserted and there were no complications. All systems responded as designed.

AOP-0007, Attachment 1, contains a graph of feedwater temperature versus core thermal power; however, no information is provided for power levels below 25.7% of rated power. With reactor power lowering to 24%, the operators were not able to determine if the unit was operating in an acceptable region. Therefore, the conservative decision was made to manually scram the reactor. Following the initial investigation, engineering input from calculations performed as part of an engineering change has provided new information for operating at power levels below 25.7%.

This event was reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) Specified System Actuation as result of Group 3 isolations (EN 56863).

This report is made pursuant to 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of the Reactor Protection System.

EVENT CAUSE

The root cause evaluation for the automatic isolation of the Low-Pressure Feedwater Heater String A is still in progress.

SAFETY ASSESSMENT

The actual consequence was the initiation of a manual scram. Following the scram, reactor pressure was maintained by the Turbine Steam Drains [TF] and reactor water level was maintained by the Feedwater [SJ] system. The RBS power to flow map continued to meet the licensing acceptance criteria parameters. There were no actual consequences to general safety of the public, nuclear safety, industrial safety, and radiological safety for this event.

CORRECTIVE ACTIONS

Clarifying information and technical basis was added to AOP-0007 to help operators better understand the applicable guidance for power operations at reduced feedwater temperatures.

The Feedwater Temperature vs Core Thermal Power Graph in AOP-0007 will be updated for power operations below 25.7%

PREVIOUS SIMILAR OCCURRENCES This section will be updated in the supplemental report following the completion of the investigation.

Energy Industry Identification System (EIIS) codes are identified in the text as [XX]. River Bend equipment codes are identified as (XX).