05000391/LER-2024-002, Re Automatic Reactor Trip Due to Steam Generator 3 Level LO-LO

Re Automatic Reactor Trip Due to Steam Generator 3 Level LO-LO

ML24127A139
Person / Time
Site:	Watts Bar
Issue date:	05/06/2024
From:	Anthony Williams Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
WBL-24-017 LER 2024-002-00
Download: ML24127A139 (1)

LER-2024-002, Re Automatic Reactor Trip Due to Steam Generator 3 Level LO-LO

Event date:
Report date:
3912024002R00 - NRC Website

text

TENNESSEE VALLEY 1\\'4 AUTHORITY

Te n n essee V alley A ut ho rity, Post Office Bo x 2000, Spring City, Tennessee 37381

WBL-24-017

May 06, 2024

10 CFR 50.73

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-391

Subject: Licensee Event Report 391 / 2024 - 002-00, Automatic Reactor Trip Due to Steam Generator #3 Level LO-LO

Pursuant to the reporting requirements of 10 CFR 50.73, attached is the subject Licensee Event Report concerning the Automatic Reactor Trip and Reactor Protection System Actuation for Watts Bar Nuc lear Plant, Unit 2, which occurred on March 05, 2024.

There are no new regulatory commitments contained in this letter. Please direct any questions concerning this matter to Jonathan Johnson, WBN Licensing Manager, at jtjohnsonO@tva.gov.

~ ~ ~-

Anthony L. Williams IV Site Vice President Watts Bar Nuclear Plant U.S. Nuclear Regulatory Commission WBL-24-017 Page 2 May 06, 2024

Enclosure : LER 391/2024-002-00, "Automatic Reactor Trip Due to Steam Generator #3 Level LO - LO "

cc (w/Enclosure) :

NRC Regional Administrator - Region 11 NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Region II ENCLOSURE Tennessee Valley Authority Watts Bar Nuclear Plant Unit 2

LER 391/2024 -002-00, " Automatic Reactor Trip Due to Steam Generator #3 Level LO -LO "

WBL-24-017 E1 of 1 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB : NO. 3150-0104 EXPIRES : 04/30/2027 (04-02-2024) Estimated

1. Facility Name rzl 050 2. Docket Number 3. Page Watts Bar Nuclea r Plant, Unit 2 052 1 OF 6 00391

4. Title Automatic Reactor Trip Due to Steam Generator #3 Level LO-LO

5. Event Date 6. LER Number 7. Report Date 8. Other Facilities Involved Month Day Year Number No. N/A 050

03 05 2024 2024 - 002 - 00 05 06 2024 Fac iN/A 052 lity Nam e Do c ket Numb er

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) 10 CFR Part 50 50. 73(a)(2)(ii)(A) 50. 73(a)(2)(viii)(A) 73.1200(a) 2O. 2201(b) 20. 2203(a)(3)(i) 50. 36(c)(1 )(i)(A) 50. 73(a)(2)(ii)(B) 50. 73(a)(2)(viii)(B) 73.1200(b) 20. 2201(d) 20.2203(a)(3)(ii) 50. 36(c)(1 )(ii)(A) 50. 73(a)(2)(iii) 50. 73(a)(2)(ix)(A) 73.1200(c) 20. 2203(a)(1) 20. 2203(a)(4) 50.36(c)(2) rzl 50. 73(a)(2)(iv)(A) 50.73( a)(2)(x) 73.1200(d) 20. 2203(a)(2)( i) 10 CFR Part 21 50.46(a)(3)(ii) 50.73(a)(2)(v)(A ) 10 CFR Part 73 73.1 200(e) 20.2203(a)(2)(ii) D 21.2(c) 50.69(9) 50.73(a)(2)(v)(B) 73. 77(a)(1) 73.1200(f) 20.2203(a)(2)(iii) 50.73(a)(2)(i)(A) 50. 73(a)(2)(v)(C) 73.77(a)(2)(i) 73.1 200(9) 20.2203(a)(2)(iv) 50.7 3(a)(2)(i)(B) 50.73(a)(2)(v)(D) 73.77(a)(2)(ii) 73.1200(h) 20.2203(a)(2)(v) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii)

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

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

K. R. Skubisz, licens ing program manager 423-365 -1771

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 A JB HS N/A y N/A N/A N/A N/A N/A

14. Supplemental Report Expected Month Day Year

15. Expected Submission Date N/A N/A N/A rzl No ID Yes ( If yes, complete 15. Ex pected S u bm ission Date

16. Abstrac t (Limit to 1326 spaces, i.e., approximately 13 single-spaced typewritten lines)

On March 03, 2024, at 0132 Eastern Standard Time (EST), with Unit 2 in Mode 1 at 100 percent power, the reacto r automat ically tripped due to a main feedwater isolation signal which resulted in a steam generator (SG) LO-LO-level reactor trip.

An aux iliary operator inadvertently manipulated 2-HS-3-945-A (Hand Switch For Control Building Isolation). This manipulation caused feedwater to be lost to all steam generators, Unit 2 tripped on LO-LO steam generator water level. A bump guard was placed over this hand-switch.

This event is being reported to the Nuclear Regulatory Commission ( NRC ) under 10 CFR 50.73(a)(2)(iv)(A) as a safety system actuation of the reactor protection system and the auxiliary feedwater (AFW) system.

I. Plant Operating Conditions before the Event

Watts Bar Unit 2 was at 100 percent Rated Thermal Power (RTP).

II. Description of Event

A. Event Summary

On Ma rch 05, 2024, at 0132 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main feedwater isolation signal which resulted in a steam generator (S G) {EIIS :SG} LO-LO-level reactor trip.

All control rods fully inserted and all automatically actuated safety related equipment operated as designed. At 0144 EST, operations personnel exited the emergency operating in structions after the plant was stabilized.

Unit 1 was not affected.

This event is being reported to the Nu clear Regulatory Commission (NRC) under 10 CFR 50.7 3(a)(2)(iv)(A) as a safety system actuation of the reactor protection system {EIIS :JC} and the auxiliary feedwater (AFW) {EIIS :BA} system.

B. Status of structures, components, or systems that were inoper able at the start of the event and that contributed to the event

There were no safety related inoperable structures, components, or systems that contributed to this event.

C. Dates and a 1proximate times of occurrences Dates and Approximate Occurrence Times An auxiliary operator inadvertently manipulates 2-HS 03/05/2024 0131 EST 945 -A. This manipulation causes feedwater to be isolated from all steam generators. Steam generator water level rapidly lowers.

Unit 2 Reactor trips on LO-LO steam generator water level.

03/05/2024 0132 EST Unit 2 enters Mode 3. Auxiliary feedwater functions r-------------+---- - --~ normally and begins to restore steam generator water level.

03/05/2024 0133 EST The operating crew entered the emergency network procedure E-0, "Reactor Trip or Safety Injection."

The operating crew exited the emergency network and 03/05/2024 0144 EST transitioned to the general operating procedure, 2-GO-5 "Unit Shutdown From 30% Reactor Power To Hot Standby."

03/05/2024 0158 EST 2-HS-3-945-A is restored to its normal pos ition.

03/05/2024 0414 EST The station notified the NRC of the event. Event number 57006 is assigned to this not ification.

D. Manufacturer and model number of each component that failed during the event

Not applicable

E. Other systems or secondary functions affected

None

F. Method of discovery of each component or system failure or procedural error

The effects resulting from the manipulation of 2-HS 945-A were immediately evident to the main control room operat ing crew. Feedwater was lost to all steam generators and 30 seconds after the switch manipulation, Unit 2 tripped on LO-LO steam generator water level.

G. The failure mode, mechanism, and effect of each failed component

Not applicable

H. Operator actions

Operations personnel promptly stabilized the plant following the reactor trip.

I. Automatically and manually initiated safety system responses

The automatic reactor trip was followed by an automatic subsequent turbine trip. Safety systems responded as expected.

Ill. Cause of the event

A. Cause of each component or system failure or personnel error

An auxiliary operator inadvertently manipulated 2-HS-3-945-A. This manipulation caused feedwater to be lost to all steam generators and 30 seconds after the switch manipulation,

Un it 2 tripped on LO - LO steam generator water level.

B. Cause(s) and circumstances for each human performance related root cause

The Human performance error reduction tool "2 minute rule" was inappropriately reduced in scope to where ingress and egress paths were not checked for hazards.

IV. Analysis of the event

While operating at approximately 100 percent power dur ing steady state operation, the Watts Bar Unit 2 reactor experienced an automatic reactor trip at approx imately 01 :32 EST on March 05,

2024, in response to a LO-LO steam generator water level. The LO-LO steam generator water level resulted from a complete loss of normal feedwater flow. Normal feedwater was lost after an auxiliary operator inadvertently manipulated 2-HS-3-945 -A. This event was compared with previous WBN plant trips and was also compared with applicable Final Safety Analys is Report (UFSAR) transients/accidents. The sequence of events associated with the trip were bounded by the FSAR Safety Analysis assumptions. The parameter response for the reactor trip is bounded by the FSAR analyses in UFSAR Section 15.2.8, " Loss of Normal Feedwater." The plant response post-trip was uncomplicated and the plant responded as designed. Operations entered E-0,

"Reactor Trip or Safety Injection," and subsequently transitioned to ES-0.1, "Reactor Trip Response," and 2-GO -5, "Unit Shutdown from 30 percent Reactor Power to Hot Standby ".

B. Corrective Actions to Prevent Recurrence or to reduce the probability of similar events occurring in the future

Master Equipment List (MEL ) changes were submitted to update the nomenclature associated with both 1-HS-3-945-A and 2-HS-3-945-A The new nomenclature states that these Hand Switches (HS) are "Fire Safe Shutdown lsol 1-FCV-3-236, 239, 242, 246 " and "Fire Safe Shutdown lsol 2-FCV-3-236, 239, 242, 246 ".

The plant label for 2-HS-3-945-A was updated with the nomenclature of "Fire Safe Shutdown lsol 2-FCV-3-236, 239, 242, 246 ".

2-HS-3-945-A was identified in the plant as a reactor trip hazard.

VII. Previous Similar Events at the Same Site

A plant trip due to a secondary plant transient which led to a loss of main feedwater was reported to the NRC in LER 391/2017-002-00 dated May 12,2017. This event was attributed to non operations personnel inadvertently depressing the local trip button for the 2A hotwell pump. The event described in this LER is different in that it involved an inadvertent equipment contact by operations personnel.

A plant trip due to a Reactor Coolant System loop low flow was reported to the NRC in LER 391 /2018-001-00 dated June 11,2018. This event was attributed to a failure to recognize the aggregate risk of minor maintenance being performed on the common drain line for the RCS Loop 1 flow transmitters. The maintenance on the common drain line was not identified as work involving trip sensitive equipment during the planning process.

VIII. Additional Information

None.

IX. Commitments

None.