05000266/LER-2021-001, Main Feedwater Pump Trip Results in Manual Reactor Trip

Main Feedwater Pump Trip Results in Manual Reactor Trip

ML21271A271
Person / Time
Site:	Point Beach
Issue date:	09/28/2021
From:	Strope M Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2021-0037 LER 2021-001-00
Download: ML21271A271 (5)

LER-2021-001, Main Feedwater Pump Trip Results in Manual Reactor Trip

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)
2662021001R00 - NRC Website

text

September 28, 2021 NRC 2021-0037 10 CFR 50.73

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

Point Beach Nuclear Plant, Unit 1 Docket 50-266 Renewed License Nos. DPR-24

Licensee Event Report 266/2021-001-00

Enclosed is Licensee Event Report (LER) 266/2021-001-00 for Point Beach Nuclear Plant, Unit 1. NextEra Energy Point Beach, LLC is providing this LER regarding the Unit 1 manual reactor trip.

This letter contains no new regulatory commitments.

If you have any questions please contact Mr. Eric Schultz, Licensing Manager, at 920-755-78~4.

Sincerely,

1.{/t~~ 1U1c.11~1u /ku,41~1r-JrJ /J;t/5

/:/~c,l~t~,(* // ( -~ /

Michael Strope Site Vice President NextEra Energy Point Beach, LLC

Enclosure

cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

NextEra Energy Point Beach, LLC

6610 Nuclear Road, Two Rivers, WI 54241 NRG FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 (08-2020) Estimated of information unless the document

• ~*~* form http_s://www.nrc.gov/reading-rm/doc-collections/nu regslstafflsr1

Point Beach Nuclear Plant Unit 1 05000266 1OF4

4. Title Main feedwater pump trip results in manual reactor trip.

5. Event Date 6. LER Number 7. Report Date 8. Other Facilities Involved Sequential Rev Facility Name Docket Number Month Day Year Year Month Day Year NA Number No. 05000NA

07 31 2021 2021 - 001 - 00 09 28 2021 Facility Name Docket Number NA 05000NA

9. Operating Mode 10. Power Level MODE1 100%

11. This Renart is Submitted Pursuant to the Reauiremenls of 1 O CFR !:l: (Check all that aoo/v) 10 CFR Part 20 D 20.2203(a)(2)(vi) D so.36(c)(2) ~ 50.73(a)(2)(iv)(A) D 50.73(a)(2)(x)

D 20.2201(b) D 20.2203(a)(3)(i) D so.46(a)(3)(ii) D 50.73(a)(2)(v)(A) 10 CFR Part 73 D 20.2201 (d) D 20.2203(a)(3)(ii) D 50.69(g) D 50.73(a)(2)(v)(B) D 73.71(a)(4)

D 20.2203(a)(1) D 20.2203(a)(4) D 50.73(a)(2)(i)(A) D 50.73(a)(2)(v)(C) D 73.71(a)(5)

D 20.2203(a)(2)(i) 1 O CFR Part 21 D 50.73(a)(2)(i)(B) D 50.73(a)(2)(v)(D) D 73.77(a)(1)(i)

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

D 20.2203(a)(2)(iii) 1 O CFR Part 50 D 50.73(a)(2)(ii)(A) D 50.73(a)(2)(viii)(A) D 73.77(a)(2)(ii)

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

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

D Other (Specify here, in Abstract, or in NRC 366A).

12. Licensee Contact for this LER Licensee Contact lihone Number (Include Area Code)

Thomas P. Schneider-Senior Licensing Engineer 920-755-7797

13. Complete One Line for each Component Failure Described in this Report Cause System Component Manufacturer Reportable To IRIS Cause System Component Manufacturer Reportable To IRIS E EA MO TECO-WH y x JI TC ABB N

14. Supplemental Report Expected Month Day Year

[g] D Yes (If 15. Expected Submission Date NA NA NA No yes, complete 15. Expected Submission Date)

Abstract

At 1646 on July 31, 2021, with Unit 1 operating in MODE 1 at full power, Operators removed Unit 1 from service by manually tripping the reactor when operators identified control board indications warranting prompt removal of the reactor from service.

All control rods fully inserted into the core due to the manual trip. The auxiliary feedwater system started as expected when a valid system actuation occurred after the reactor trip. There was no emergency core cooling system actuation and offsite power was maintained throughout the event. After the reactor trip, equipment anomalies occurred for which the operators effectively responded.

The cause of the condition requiring the manual reactor trip was a failure of the main steam generator feedwater pump B motor. Corrective actions included replacement of the main steam generator feedwater pump B motor and additional maintenance activities for anomalies identified during the reactor trip.

This event is reportable in accordance with 10 CFR 50. 73(a)(2)(iv)(A).

Description of the Event:

At 1646 on July 31, 2021, with Unit 1 operating in MODE 1 at full power, Operators removed Unit 1 from service by manually tripping the reactor [AC] '<<hen operators identified control board indications warranting prompt removal of the reactor from service.

All control rods [JD] fully inserted into the core due to the manual trip. The auxiliary feedwater system [BA] started as expected when a valid system actuation occurred after the reactor trip. There was no emergency core cooling system actuation [JE] and offsite power [FK] was maintained throughout the event.

After the reactor trip, a condenser steam dump valve [JI] cycled and did not fully close. The valve was locally isolated to prevent additional reactor [AC] cooldown. After the turbine trip, the crossover steam dump system motor operated valves

[SE] did not close. This caused main condenser [SG] vacuum conditions to deteriorate. This lead to main condenser unavailability and the use of the atmospheric steam dump system. Field action was taken to close the crossover steam dump valves. Additionally, during the feedwater transition, the main feedwater B regulating bypass valve [JB] did not control in automatic and was taken to manual control.

• This 60-day licensee event report is being submitted in accordance with the requirements of 10 CFR 50. 73(a)(2)(iv)(A).

Cause of the Event

The cause of the manual reactor trip was due to the failure of main steam generator feedwater (SGFW) pump B motor

[EA][MO]. The root cause has been determined to be less than adequate manufacturing process controls for the stator and rotor assembly. A misalignment between the stator and rotor resulted in periodic contact of the rotor to stator eventually degrading the iron laminations and winding insulation to the point of failure of the stator winding.

The condenser steam dump valve [JI] condition was caused by the failure of its positioner [TC]. The crossover steam dump valve [SE] condition was caused by the failure of a relay [RL Y], and the feedwater regulating valve condition will be investigated further during the next available opportunity.

Analysis of the Event

The feedwater system is comprised of two half-capacity main SGFW pumps that deliver fluid to the secondary side of the steam generators. The fluid is used to remove the heat generated in the reactor and to produce steam used for electrical power generation.

The loss of the main SGFW pump motor required an' immediate reduction in reactor power to support the limitations of a single train of feedwater. At the time of the failure the reactor power was not low enough to support removal of one of the two operating main SGFW pumps, which necessitated a manual reactor trip.

The Auxiliary Feedwater Pumps started as expected on low steam generator level experienced due to reducing steam demand from the turbine trip in response to the reactor trip.

After the reactor trip, all control rods fully inserted in the core due to the manual trip. There was no Emergency Core Cooling System actuation. Offsite power was maintained throughout the event. The equipment anomalies that occurred did not affect safety system functions.

Corrective Actions

The anomalies experienced during the shutdown were diagnosed, and those requiring maintenance prior to reactor restart were completed. The reactor was returned to service at reduced power and the main SGFW pump motor was replaced NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 (08-2020)

Point Beach Nuclear Plant Unit 1 05000266 NUMBER NO.

2021 -,001 - 00 prior to returning the reactor to full power. Additionally, procurement specifications will be updated to include industry approved standards and oversight criteria for critical motor assembly.

Safety Significance

During the event and subsequent recovery actions, there was no loss of any safety systems, structures, or components.

The auxiliary feedwater pumps started as expected during the transient. The main SGFW pump A remained available to remove decay heat after the reactor trip. Plant systems anomalies occurred following the manual reactor trip. The operating crew appropriately responded to the anomalies. Following the manual reactor trip, all control rods fully inserted into the core as designed to control reactivity and temperature of the core. The reactivity effects during this event had no impact on the safety of the core and thus, the event was determined to be of very low safety significance. There was no impact on the health and safety of the public because of this event.

Similar Events

There have not been similar events of manual reactor trips in the past three years with the same initiating condition.

Component Failure Data

Main Steam Generator Feedwater Pump Motor: TECO Westinghouse - Frame 7118 Main Steam Condenser Dump Valves: ABB Inc-Model AV232300 Crossover Steam Dump Valves: Cutler Hammer - Model BFD20S NRC FORM 366B U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 4/30/2020 (08-2020)

Point Beach Nuclear Plant Unit 1 05000266 NUMBER NO.

2021 - 001 - 00

13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO IRIS x SE RLY Cutler Hammer -C770 N

I

NRG FORM 3668 (08-2020) Page 4 of 4