05000416/LER-2022-003, Manual Reactor Scram Due to Trip of the a Reactor Feedwater Pump

Manual Reactor Scram Due to Trip of the a Reactor Feedwater Pump

ML23047A547
Person / Time
Site:	Grand Gulf
Issue date:	02/16/2023
From:	Hardy J Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
GNRO2023-00006 LER 2022-003-00
Download: ML23047A547 (1)

LER-2022-003, Manual Reactor Scram Due to Trip of the a Reactor Feedwater Pump

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv)(B), System Actuation
4162022003R00 - NRC Website

text

Entergy Operations, Inc.

entergy P.O. Box756 Port Gibson, Mississippi 39150

Jeffery A. Hardy Manager Regulatory Assurance Grand Gulf Nuclear Station Tel: 802-380-5124

10 CFR 50.73

G N RO2023-00006

February 16, 2023

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

SUBJECT: Grand Gulf Nuclear Station, Unit 1 Licensee Event Report 2022-003-00, Manual Reactor SCRAM due to trip of the A Reactor Feedwater Pump

Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 Renewed License No. NPF-29

Attached is Licensee Event Report 2022-003-00, Manual Reactor SCRAM due to trip of the A Reactor Feedwater Pump. This report is being submitted in accordance with 1 0 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of the Reactor Protection System.

This letter contains no new Regulatory Commitments. Should you have any questions concerning the content of this letter, please contact Jeff Hardy, Regulatory Assurance Manager at 802-380-5124.

Sincerely, 9A~

JAH/saw

Attachments: Licensee Event Report 2022-003-00 G N RO2023-00006 Page 2 of 3

cc: NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

U.S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 G N RO2023-00006 Page 3 of 3

Attachment Licensee Event Report 2022-003-00

Abstract

At 2058 on December 19, 2022, Grand Gulf Nuclear Station, Unit 1 experienced a trip of Condensate Booster Pump A.

At 2101, Reactor Feedwater Pump A tripped, and the reactor was manually scrammed. All control rods were fully inserted into the core. At 2104, Reactor Feedwater Pump B tripped. At 2121 High Pressure Core Spray (HPCS) was manually injected to maintain reactor water level. At 2126 Reactor Feedwater Pump A was successfully restarted.

Following the reactor SCRAM, the HPCS system was used to maintain reactor water level. The direct cause of the SCRAM was the Heater Drain Tank (HOT) Level Control Valves were not capable of responding quickly enough to mitigate the HOT level oscillations during the transient caused by Condensate Booster Pump Trip and subsequent Fast Detent.

There were no consequences to the safety of the public, nuclear safety, industrial safety, or radiological safety. No radiological releases occurred due to this event. This report is made in accordance with 10 CFR 50.73(a)(2)(iv)(A) for any event or condition that resulted in manual of automatic actuation of any system listed in paragraph 10 CFR 50.73 (a)(2)(iv)(B). The Reactor Protection System including reactor SCRAM or trip is included in 10 CFR 50.73 (a)(2)(iv)(B).

PLANT CONDITIONS

Mode 1, 78 percent power

DESCRIPTION OF EVENTS At 2058 on December 19, 2022, while the plant was operating at 100 percent reactor power, Condensate Booster Pump (CBP)[SD] A tripped. Control Room Operators entered the Feedwater System Malfunction Off-Normal Event Procedure (ONEP) and reduced core flow to 70 million pound-mass per hour. While performing steps of the ONEP operators identified Reactor Feedwater Pump "A" and "B" suction pressures were rapidly lowering. At 2101, Reactor Feedwater Pump "A" tripped on low suction pressure. A recirc flow control valve runback occurred as designed. Reactor water level continued to lower because of a reduction in feed flow. At approximately 17 inches of water and trending down, the At-The-Controls Operator placed the Reactor Mode Switch in Shutdown at 2101. At 2121 High Pressure Core Spray (HPCS) [BG] was manually injected to maintain reactor water level.

The manual SCRAM on low reactor water level resulted from constricting Feed Flow when the Heater Drain Tank Level Control Valves closed, which is a plant response that should have been found in the results of calculation for the "Condensate and Feedwater Transient Analysis" (CFT A). This oversight, along with other issues found in the CFT A calculation led to the Root Cause of "Transients caused by CBP Trip have not been adequately analyzed", and this condition will be corrected by performing an analysis to determine the best corrective modification and implementing that modification during the next refueling outage.

REPORT ABILITY This report is made in accordance with 10 CFR 50.73(a)(2)(iv)(A) for any event or condition that resulted in manual of automatic actuation of any system listed in paragraph 10 CFR 50. 73 (a)(2)(iv)(B). The Reactor Protection System including reactor SCRAM or trip is included in 10 CFR 50.73 (a)(2)(iv)(B).

CAUSE

Direct Cause: Heater Drain Tank (HOT) Level Control Valves were not capable of responding quickly enough to mitigate the HOT level oscillations during the transient caused by Condensate Booster Pump Trip and subsequent Fast Detent.

Root Cause: Transients Caused by Condensate Booster Pump Trip have not been adequately analyzed.

CORRECTIVE ACTIONS

1. Replace CBP A Motor - The CBP A Motor has been replaced.

2. Corrective Action to Prevent Recurrence (CAPR) - Modification will be performed to implement long term solutions as determined to analyze the transient effects of the following on Feed Flow and Post Extended Power Uprate (EPU) and Maximum Extended Load Line Limit Analysis Plus (MELLA+) conditions: Condensate Pump trip, CBP trip, HOP trip. Develop potential solutions to solve latent design issues in Feed Flow transient response.NRC FORM 366A APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 12/31/2023 (01-10-2023) U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME [81 050 2. DOCKET NUMBER 3. LER NUMBER Grand Gulf Nuclear Station, Unit 1 416 NUMBER NO. 052 2022 - 003 - 00

SAFETY SIGNIFICANCE

There were no consequences to the safety of the public, nuclear safety, industrial safety, or radiological safety. No radiological releases occurred due to this event.

Following the SCRAM, HPCS remained operable and was used to feed the reactor. Safety/ Relief Valves were operable and available to reduce reactor pressure if required and safety-related Emergency Core Cooling Systems were operable.

The condensate and feedwater systems are described in the Section 10.4.7 of the Updated Final Safety Analysis Report (UFSAR). The function of the condensate and feedwater system is to provide a dependable supply of feedwater to the reactor, to provide feedwater heating, and to maintain high water quality in the feedwater. The system is designed to provide the required flow at the required pressure to the reactor, allowing sufficient margin to provide continued flow under anticipated transient conditions.

The condensate and feedwater system serves no safety function. Systems analysis has shown that failure of this system will not compromise any safety-related systems or prevent safe shutdown. The condensate and feedwater system is not required to effect or support the safe shutdown of the reactor or perform in the operation of reactor safety features.

PREVIOUSLY SIMILAR EVENTS A review of internal operating experience for the previous three years identified the following similar event.

LER 2020-004-01 Automatic Reactor SCRAM Due to Reactor Feed Pump Trip