Event Notification Report for December 13, 2019

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
12/12/2019 - 12/13/2019

** EVENT NUMBERS **

Power Reactor	Event Number: 54438
Facility: SEQUOYAH Region: 2 State: TN Unit: [] [2] [] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: FRANCIS SCHULTE HQ OPS Officer: JEFF HERRERA	Notification Date: 12/12/2019 Notification Time: 08:14 [ET] Event Date: 12/12/2019 Event Time: 04:33 [EST] Last Update Date: 12/12/2019
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION	Person (Organization): BRIAN BONSER (R2DO)

Unit

SCRAM Code

RX Crit

Initial PWR

Initial RX Mode

Current PWR

Current RX Mode

M/R

100

Power Operation

Hot Standby

Event Text

MANUAL REACTOR TRIP DUE TO A LOSS OF HEATER DRAIN TANK PUMP FLOW

"At 0432 EST, on 12/12/19, Sequoyah Unit 2 experienced a manual reactor trip. The trip was initiated due to a loss all number 3 Feedwater Heater Drain Tank pump flow; plant procedures directed a manual reactor trip if power is greater than 80 percent.

"The Auxiliary Feedwater System (AFW) automatically actuated as required when the expected post trip feedwater isolation actuation actuated. Reactor Coolant System (RCS) temperature is being maintained by the steam dump system with all 4 Reactor Coolant Pumps (RCPs) in service.

"All control and shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the transient. Unit 2 is currently stable at normal operating temperature and normal operating pressure in Mode 3. The electrical system is in a normal alignment.

"There was no impact on U1. There was no impact to the health and safety of the public or plant personnel.

"Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four hour, non-emergency notification per 10CFR50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification accordance with 10CFR50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system."

The NRC Resident Inspector was notified.

Part 21	Event Number: 54440
Rep Org: WESTINGHOUSE Licensee: WESTINGHOUSE Region: 1 City: CRANBERRY TOWNSHIP State: PA County: License #: Agreement: Y Docket: NRC Notified By: CAMILLE T. ZOZULA HQ OPS Officer: OSSY FONT	Notification Date: 12/12/2019 Notification Time: 19:24 [ET] Event Date: 12/12/2019 Event Time: 00:00 [EST] Last Update Date: 12/12/2019
Emergency Class: NON EMERGENCY 10 CFR Section: 21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE	Person (Organization): CHRISTOPHER LALLY (R1DO) BRIAN BONSER (R2DO) BILLY DICKSON (R3DO) JEREMY GROOM (R4DO) - PART 21/50.55 REACTORS (EMAIL)

Event Text

PART 21 REPORT - FRACTURED AND DISLOCATED CONTROL ROD DRIVE MECHANISM THERMAL SLEEVE

"During a 2019 planned outage at a Westinghouse plant, site personnel identified a fractured and dislocated control rod drive mechanism (CRDM) thermal sleeve. The fracture occurred just beneath the worn area of the flange in the full cross-section of the thermal sleeve tube. A stress concentration exists at this transition. Previous operating experience (OE) with thermal sleeve failures did not include a cross-sectional thermal sleeve fracture such as this.

"Additional data supplied from the affected plant showed evidence of additional thermal sleeve locations with crack-like indications in the flange collar region (i.e., evidence of degradation but not failure). Although there was no evidence that control rod motion was hindered, Westinghouse is conservatively reporting this condition as having the potential to create a SSH (substantial safety hazard), were it to remain uncorrected.

"Based on new OE provided to Westinghouse, a defect has been identified that is associated with a previously unseen form of thermal sleeve degradation (i.e., mechanical fatigue and fracture that leads to flange separation). Control rod functionality could become adversely impacted not only due to the flange wear reported in LTR-NRC-18-34, but due to the additional coincident fracture and separation of the thermal sleeve tube from its flange. This condition could exist prior to reaching the flange wear criteria established in PWROG-16003-P, Revision 2. The information supplied in PWROG-16003-P, Revision 2 and NSAL-18-1 also does not address this new OE. If no action is taken to monitor and/or correct this condition, an SSH could occur if the insertion of more than one control rod is prevented.

"The probability for this to result in a SSH is low given that this is the very first observance of this phenomenon. Westinghouse does not expect that an affected plant would experience two or more stuck control rods during its current operating cycle. Even if multiple stuck control rods were to occur, such an event would be bounded by the licensee's ATWS analysis. Based on known wear conditions and wear rates, plants can safely operate for at least one cycle or until the next opportunity to perform a visual inspection.

"The potentially affected plants listed below are Westinghouse design plants that:
1. Operate with higher upper head bypass flow conditions, known as 'T-cold' head plants, and
2. Operate with thermal sleeves containing a collar (or upper centering pad ring) just below the flange.
Asco 2, Braidwood 1, Braidwood 2, Byron 1, Byron 2, Callaway 1, Catawba 1, Catawba 2, Comanche Peak 1, Comanche Peak 2, Doel 4, Hanbit 1, Hanbit 2, Kori 3, Kori 4, Maanshan 1, Maanshan 2, Seabrook, Sizewell B, Tihange 3, Vogtle 1, Vogtle 2, and Wolf Creek.

"A Westinghouse communication will be supplied to affected licensees in early 2020 to inform them that this defect has been reported. The communication will provide updated recommendations concerning future inspection guidance."

The person informing the Nuclear Regulatory Commission:
Camille T. Zozula
Westinghouse Electric Company
1000 Westinghouse Drive
Cranberry Township, Pennsylvania 16066
Direct tel: (412) 374-2577
Direct fax: (724) 940-8542
e-mail: zozulact@westinghouse.com

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