05000335/LER-2021-001, Manual Reactor Trip Due to Insufficient Feed Flow to Steam Generators

Manual Reactor Trip Due to Insufficient Feed Flow to Steam Generators

ML22035A033
Person / Time
Site:	Saint Lucie
Issue date:	02/04/2022
From:	Deboer D Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L -2022-019 LER 2021-001-00
Download: ML22035A033 (4)

LER-2021-001, Manual Reactor Trip Due to Insufficient Feed Flow to Steam Generators

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A)
3352021001R00 - NRC Website

text

F=PL.. February 4, 2022 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Re: St. Lucie Unit 1 Docket No. 50-335 Reportable Event: 2021-001-00 Date of Event: December 10, 2021 L-2022-019 10CFR 50.73 Manual Reactor Trip Due to Insufficient Feed Flow to Steam Generators The attached Licensee Event Report 2021-001 is being submitted pursuant to the requirements of 10 CFR 50.73 to provide notification of the subject event.

Respectfully, Daniel DeBoer Site Vice President St. Lucie Plant DD/tf Attachment cc: St. Lucie NRC Senior Resident Inspector St. Lucie NRC Program Manager Florida Power & Light Company 6501 S. Ocean Drive, Jensen Beach, FL 34957 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 (08-2020)

St. LucieUnit 1 050003 35 1 of 3

4. Title Manual Reactor Trip Due to Insufficient Feed Flow to Steam Generators

5. Event Date 6. LER Number 7. Report Date 8. Other Facilities Involved Revision Facility Name Docket Number Month Day Year Year Sequential Number Number Month Day Year n/a 05000

Facility Name Docket Number 12 10 2021 2021 - 001 - 00 02 04 2022 n/a 05000

9. Operating Mode 10. Power Level 1 100

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 10 CFR Part 20 20.2203(a)(2)(vi) 50.36(c)(2) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x)

20.2201(b) 20.2203(a)(3)(i) 50.46(a)(3)(ii) 50.73(a)(2)(v)(A) 10 CFR Part 73

20.2201(d) 20.2203(a)(3)(ii) 50.69(g) 50.73(a)(2)(v)(B) 73.71(a)(4)

20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C) 73.71(a)(5) 20.2203(a)(2)(i) 10 CFR Part 21 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D) 73.77(a)(1)

20.2203(a)(2)(ii) 21.2(c) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 73.77(a)(2)(i)

20.2203(a)(2)(iii) 10 CFR Part 50 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 73.77(a)(2)(ii)

20.2203(a)(2)(iv) 50.36(c)(1)(i)(A) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B)

20.2203(a)(2)(v) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)

OTHER (Specify here, in abstract or in NRC 366A)

12. Licensee Contact for this LER Licensee Contact Telephone Number (Include area code)

Timothy Falkiewic] Licensing Engineer ( ) - 56

13. Complete One Line for each Component Failure Described in this Report Cause System Component Manuf acturer Reportable Cause System Component Manuf acturer Reportable To IRIS To IRIS A SN PDIS I204 Y

14. Supplemental Report Expected 15. Expected Month Day Year YES (If yes, complete 15. Expected Submission Date) NO Submission Date Abstract (Limit to 156 0 spaces, i.e., approximately 15 single-spaced typewritten lines)

On December 10, 2021 at 1024 EST with Unit 1 in Mode 1 at 100% power, the reactor was manually tripped due to lowering level in the steam generators. The trip was uncomplicated with all systems responding normally post trip.

Operators restored steam generator level utilizing main feedwater and stabilized the reactor in Mode 3.

The event occurred while replacing a feedwater heater control Pressure Differential Indicating Switch (PDIS). In the process of landing the wires from the new PDIS on the terminal strip, the technician made inadvertent contact with the enclosure housing resulting in a blown supply fuse which causeda loss of high pressure heater level control and a reduction of Steam Generator feedwater flow.

Due to the emergent nature of this work, some of the normal preplanned mitigation measures were not in place.

Corrective actions include maintenance procedure revisions.

Manual reactor trips are analyzed events in the Updated Final Safety Analysis Report. The trip was uncomplicated.

Therefore, this event had no impact on the health and safety of the public.

Description

On December 10, 2021 at 1024 EST with Unit 1 in Mode 1 at 100% power, the reactor was manually tripped in response to Steam Generator Water Level lowering less than 50% from a secondary transient initiated from the loss of the feedwater heater level control system [SN]. The trip was uncomplicated with all systems responding normally post trip. Operators restored steam generator level utilizing main feedwater [SJ] and stabilized the reactor in Mode 3.

The event occurred while replacing a feedwater heater control Pressure Differential Indicating Switch [SN:PDIS].

In the process of landing the wires from the new PDIS on the terminal strip, the technician made inadvertent contact with the enclosure housing resulting in a blownsupply fuse [SN:FU] and a loss of Moisture Separator Heater [MSR] drain collector 1C level control which causeda reductionof Steam Generator feedwater flow.

Cause of the Event

Due to the emergent nature of this work, some of the normal preplanned mitigation measures were not in place.

A Maintenance team was replacing the Unit 1 PDIS-11-30D1, Pressure Differential Indicating Switch Between High Pressure Heater 5A and Moisture Separator Reheater 1D, due to a steam leak on the PDIS. At the time the 1D MSR drain collector level control valve was placed in manual control to maintain MSR reheater level.

During electrical connection of the new instrument, the technician inadvertently touched the wire to the enclosure. This resulted in a blown power supply fuse and closure of the 1C MSR drain collector level control valve, which caused a reduction of Steam Generator feed flow. The reactor was tripped due to lowering and unrecoverable level in the Steam Generators.

Analysis of the Event

This licensee event report is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B).

This event included manual actuation of the Reactor Protection System (RPS). This event had no significant safety consequence since the RPS successfully performed its intended safety function.

Safety Significance

Reactor trip events are described in the UFSAR as anticipated operational occurrences.

There were no complications, and all safety related systems functioned as designed. There were no safety system actuations other than RPS as a result of the event. Given the response of the plant and the plant design that can accommodate this anticipated operational occurrence, the health and safety of the public were not affected by this event.

Corrective Actions

The corrective actions include revising maintenance control procedures to add Senior Reactor Operator reviews for certain emergent work.

Failed Components Identified

1) PDIS 30D1

Description

PDIS between the HP heater 5A and the 1D MSR

Part Number: 288A (0- 100 psig)

Manufacturer: ITT - Barton Instrument

Similar Events

None