05000316/LER-2024-002, Manual Reactor Trip Following Rapid Downpower for Steam Leak

Manual Reactor Trip Following Rapid Downpower for Steam Leak

ML24197A088
Person / Time
Site:	Cook
Issue date:	07/15/2024
From:	Ferneau K Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
AEP-NRC-2024-52 LER 2024-002-00
Download: ML24197A088 (1)

LER-2024-002, Manual Reactor Trip Following Rapid Downpower for Steam Leak

Event date:
Report date:
3162024002R00 - NRC Website

text

s INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER One Cook Place Bridgman, Ml 49106 A unit of American Electric Power lndianaMichiganPower.com

July 15, 2024 AEP-NRC-2024-52 10 CFR 50.73 Docket No.: 50-316

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Donald C. Cook Nuclear Plant Unit 2 LICENSEE EVENT REPORT 316/2024-002-00 Unit 2 Manual Reactor Trip Following Rapid Downpower for Steam Leak

In accordance with 10 CFR 50. 73, Licensee Event Report (LER) System, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant Unit 2, is submitting as an enclosure to this letter the following report:

LER 316/2024-002-00: Unit 2 Manual Reactor Trip Following Rapid Downpower for Steam Leak

There are no commitments contained in this submittal.

Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely,

Kelly J. Ferneau Site Vice President

MPH/sjh

Enclosure: Licensee Event Report 316/2024-002-00: Unit 2 Manual Reactor Trip Following Rapid Down power for Steam Leak

c : EGLE - RMD/RPS J. B. Giessner - NRC Region Ill NRC Resident Inspector N. Quilico - MPSC R. M. Sistevaris -AEP Ft. Wayne S. P. Wall, NRC WasJ,ington D.C.

A. J. Williamson -AEPJ Ft. Wayne Enclosure to AEP-NRC-2024-52

Licensee Event Report 316/2024-002-00: Unit 2 Manual Reactor Trip Following Rapid Downpower for Steam Leak

Abstract

On May 15, 2024, at 04:27 EDT, the Donald C. Cook Nuclear Plant (CNP) Unit 2 reactor was manually tripped from 20%

reactor power following a rapid downpower in response to a steam leak. All control rods fully inserted, and electrical power was supplied by offsite. Auxiliary Feedwater Pumps automatically started as required, and decay heat removal was through the Condenser Steam Dump system. All required equipment operated as expected, and the trip was not complicated.

A leak on a one-inch diameter steam line required a unit downpower to take the Main Turbine offline, which led to diverging steam generator level oscillations that resulted in operators manually tripping the reactor. The j eak was repaired, and operators restarted the unit normally. A supplemental report is expected based on the conclusions of an ongoing root cause evaluation.

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The event is reportable in cjlCCordance with 10 CFR 5O.73(a)(2)(iv)(A), System A 9tuation, due to the valid actuation of the l Reactor Protection System and the Auxiliary Feedwater System, as a result of the manual reactor trip.

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EVENT DESCRIPTION

On May 15, 2024, at 04:27 EDT, the Donald C. Cook Nuclear Plant {CNP) Unit 2 Reactor was manually tripped from 20 percent power following a rapid downpower in response to a steam leak on a one-inch diamater steam line.

Unit 2 was supplied by offsite power before and after the reactor trip. All control rods fully inserted. Following the trip, the

~uxiliary Feedwater {AFW)[BA] Pumps [P] started as required and operated as designed. Decay heat removal was through

~he Condenser Steam Dump system [Jl][COND]. All required equipment operated as expected, and the trip was not complicated.

Prior to the event, CNP performed a downpower of Unit 2 to take the Main Turbine [TAJ offline for repair of a leak on a one-inch diameter steam line downstream of the Right Outer High Pressure Turbine Stop-Control Valve [XCV]. During the downpower, the leak worsened and operators transitioned to the rapid downpower procedure. Operators manually tripped the Main Turbine from 25 percent power at 03:55 EDT to isolate the steam leak. Following the turbine trip, Steam Generator

[SB][SG] water level oscillations resulted in operators manually tripping the reactor.

Event Notification 57128 was submitted in accordance with 10 CFR 50.72{b){2){iv){B), Reactor Protection System

{RPS) actuation as a four (4) hour non-emergency report, and under 10 CFR 50.72{b){3){iv){A), specified system actuation of the AFW System, as an eight (8) hour non-emergency report.

The event is being reported in accordance with 10 CFR 50.73{a){2){iv){A), System Actuation, due to the valid actuation of the RPS and the AFW System, as a result of the manual reactor trip.

CAUSE OF THE EVENT

The direct cause of the event was the leak on a one-inch diameter steam line downstream of the Right Outer High Pressure Turbine Stop-Control Valve that worsened during the downpower. Operators transitioned to the rapid downpower procedure to expedite tripping the Main Turbine to isolate the steam leak. Following the Main Turbine trip, Steam Generator water level oscillations resulted in operators manually tripping the reactor prior to reaching pre-established operational limits.

A Root Cause Evaluation {RCE) is in progress at the time of the writing of this Licensee Event Report {LER). After the approval of the RCE, a supplement report to this LER will be submitted regarding the cause of the event.

IMMEDIATE CORRECTIVE ACTIONS

The steam leak downstream of the Right Outer High Pressure Turbine Stop-Control Valve was repaired.

PLANNED CORRECTIVE ACTIONS

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Corre6tive actions resulting from the RCE will be included in the supplement report to this LER.

ASSESSMENT OF SAFETY CONSEQUENCES

NUCLEAR SAFETY

All required equipment operated as designed and there was no actual or potential nuclear safety hazard resulting from the steam generator water level oscillations and subsequent reactor trip.

INDUSTRIAL SAFETY

There was no actual personnel safety hazard resulting from the steam generator water level oscillations and subsequent reactor trip.

RADIOLOGICAL SAFETY

There was no actual or potential radiological safety hazard, or radiological release, resulting from the steam generator water level oscillations and subsequent reactor trip.

PROBABILISTIC RISK ASSESSMENT (PRA)

IA PRA risk assessment was performed by calculating the Conditional Core Damage Probability (CCDP) and the Conditional Large Early Release Probability (CLERP) of the transient initiating event. A comparison of the results of these calculations to the thresholds provided in NRC Inspection Manual Chapter (IMC) 0609, determined this event to be of "Very Low Safety Significance".

PREVIOUS SIMILAR EVENTS

LER 316/2022-001-00, Automatic Reactor Trip Due to High-High Steam Generator Level, was submitted to the NRC as an enclosure to letter AEP-NRC-2023-01 (ML23004A205). The scope of the RCE conducted for the previous event was specific to operating conditions of the event and did not include low power operating conditions with the main turbine offline following a rapid downpower.

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