05000387/LER-2023-004-01, Manual Reactor Scram Due to Degraded Main Condenser Vacuum

Manual Reactor Scram Due to Degraded Main Condenser Vacuum

ML24239A847
Person / Time
Site:	Susquehanna
Issue date:	08/26/2024
From:	Casulli E Susquehanna, Talen Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
PLA-8139 LER 2023-004-01
Download: ML24239A847 (1)

LER-2023-004, Manual Reactor Scram Due to Degraded Main Condenser Vacuum

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation
3872023004R01 - NRC Website

text

Edward Casulli Site Vice President Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3795 Fax 570.542.1504 Edward.Casulli@TalenEnergy.com SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-387/2023-004-01 UNIT 1 LICENSE NO. NPF-14 PLA-8139 TALEN~

ENERGY 10 CFR 50.73 Docket No. 50-387 Attached is Licensee Event Report (LER) 50-387/2023-004-01. The LER supplement reports an event involving a manual scram due to degrading main condenser vacuum. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the Reactor Protection System (including a reactor scram).

There were no actual consequences to the health and safety of the public as a result of this event.

This letter contains no new or revised regulatory commitments.

E. Casulli Attachment: LER 50-387/2023-004-01 Copy:

NRC Region I Ms. J. England, NRC Senior Resident Inspector Ms. A. Klett, NRC Project Manager Mr. M. Shields, PADEP/BRP August 26, 2024

Abstract

On November 10, 2023, at approximately 01: 18, Susquehanna Steam Electric Station Unit 1 reactor was manually scrammed due to degrading Main Condenser vacuum caused by a failed turbine bearing waste water and oil (slop) drain. The event was reported by Event Notification 56846 in accordance with 10 CFR 50. 72(b )(2)(iv)(B) and 10 CFR 50. 72(b )(3)(iv)(A). This event is also reportable in accordance with 10 CFR 50. 73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the Reactor Protection System (including a reactor scram).

Failure of the slop drain was caused by vibration/cyclic fatigue-induced failure of the socket weld on the 1.5" pipe to the 3" x 1.5" pipe reducer inside the Main Condenser. Key corrective action included plugging and welding the failed slop drain and reducing the length of the resulting pipe stub.

There were no actual consequences to the health and safety of the public as a result of this event.

CONDITIONS PRIOR TO EVENT

2. DOCKET NUMBER

3. LER NUMBER I

00387 r:::J NUMBER NO.

I YEAR SEQUENTIAL REV

~-I 004 1-0 Unit 1 - Mode 1, approximately 100% Rated Thermal Power (RTP)

Unit 2 - Mode 1, approximately 100% RTP Vacuum rapidly degraded on the Susquehanna Steam Electric Station (SSES) Unit 1 Main Condenser, which resulted in the need for a manual reactor scram.

EVENT DESCRIPTION On November 10, 2023, SSES Unit 1 was manually scrammed due to degraded Main Condenser [EIIS System Code/

Component Code: SG/COND] vacuum caused by a failed turbine bearing waste water and oil (slop) drain [EIIS System Code/Component Code: TF/DRN]. The following is a timeline of significant events associated with the scram:

November 10, 2023, at approximately 01 :06 - Unit 1 Main Condenser vacuum was rapidly degrading along with indication of high offgas flow. A Recirculation Limiter 2 runback was inserted to lower reactor power. Main Condenser vacuum continued to degrade following the reduction in reactor power to approximately 70%.

November 10, 2023, at approximately 01: 18 - A manual scram was inserted when Main Condenser vacuum reached 6 inches of mercury absolute. Plant response to the scram was per design, Reactor Protection System (RPS) [EIIS System Code: JC] channels all tripped and maintained the scram signal for the required 10 second period. Reactor Water Level lowered to -25" following the scram. Reactor Water Level 3 (+13 inch) containment isolation system signals were received and went to completion. The feedwater system [EIIS System Code: SJ] remained in service as the primary Reactor Pressure Vessel (EIIS System Code/Component Code: AC/RPV] water level control system. There were no Emergency Diesel Generator [EIIS System Code/Component Code: EK/DG] starts or Safety Relief Valve [EIIS System Code/

Component Code: SB/RV] actuations during the event.

The event was reported by Event Notification 56846 in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)

(iv)(A). This event is also reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the RPS (including a reactor scram).

CAUSE OF EVENT Failure of the slop drain was caused by vibration/cyclic fatigue-induced failure of the socket weld on the 1.5" pipe to the 3" x 1.5" pipe reducer inside the Main Condenser. This failure was due to the original slop drain capping and piping removal modification in 2002 not being performed in accordance with the associated engineering change. In addition to the slop drain piping, pipe supports that were intended to support the remaining piping were erroneously removed. This allowed a remaining vertically suspended 1.5" diameter pipe stub of approximately 40 inches to vibrate, which eventually caused the weld failure.

ANALYSIS/SAFETY SIGNIFICANCE The actual consequences from the drain leaking air into the Unit 1 Main Condenser were degraded Unit 1 Main Condenser vacuum, a forced down power, and subsequent manual scram.

Without prompt operator action, an automatic scram would have been initiated. Additionally, High Pressure Coolant Injection (EIIS System Code: BJ) System and Reactor Core Isolation Cooling (EIIS System Code: BN) System could have initiated which would require initiating support systems as well as adding additional heat to the suppression pool. This would have necessitated using the Residual Heat Removal (EIIS System Code: BJ) System in Suppression Pool Cooling mode to transfer the suppression pool heat to the spray pond.

2. DOCKET NUMBER

3. LER NUMBER I

00387 IB-1 SEQUENTIAL REV NUMBER NO.

004 1-0 Based on the results of a risk significance evaluation, this event is classified as having "very low" safety significance. It should also be noted that this manual scram was uncomplicated, did not require Emergency Core Cooling System initiation, and Main Steam Isolation Valves remained open (i.e. the main condenser remained available as the primary heat sink).

The condition described herein did not result in a Safety System Functional Failure. Accordingly, this event will not be counted as a Safety System Functional Failure in the Reactor Oversight Process Performance Indicators. There were no actual consequences to the health and safety of the public as a result of this event.

CORRECTIVE ACTIONS Following the scram, a gasketed plate and plug assembly was installed at the failed slop drain, allowing Unit 1 to resume normal operation until permanent repairs could be made. This temporary engineering change was also preemptively performed on other Unit 1 slop drain piping stubs prior to resuming normal operation.

Dtiring the 2024 refueling and inspection outage, the failed slop drain piping was plugged and welded, and the length of the remaining 3" diameter pipe stub reduced. As part of the extent of condition, the temporary engineering change described above may be installed on the remaining affected slop drains until permanent repairs are complet~.

COMPONENT FAILURE INFORMATION Component Identification - Weld at 3" to 1.5" reducer Component Name - Waste Water and Oil Drain piping Component Part Number - N/A Manufacturer - General Electric (GE)

PREVIOUS OCCURRENCES None. Page 3

of 3