05000341/LER-2022-001, Reactor Scram on Low Reactor Pressure Vessel Level

Reactor Scram on Low Reactor Pressure Vessel Level

ML22094A155
Person / Time
Site:	Fermi
Issue date:	04/04/2022
From:	Peter Dietrich DTE Electric Company
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-22-0013 LER 2022-001-00
Download: ML22094A155 (5)

LER-2022-001, Reactor Scram on Low Reactor Pressure Vessel Level

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation
3412022001R00 - NRC Website

text

Pdcr Didrich Se nior Vice President and C hief Nucle:u Officer

DT E E lectric C ompan y 6400 N. Dixie Highway, Newport, Ml 48166 Tel: 73 4. 58 6.41 53 Email: petcr.dictrich @ dtccner gy.co111 DTE

April 4, 2022 10 CFR 50.73 NRC-22-0013

U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Fermi 2 Power Plant NRC Docket No. 50-341 NRC License No. NPF-43

Subject: Licensee Event Repo1i (LER) No. 2022-001

Pursuant to 10CFR50.73(a)(2)(iv)(A), DTE Electric Company (DTE) is submitting LER No.

2022-00 l, Reactor Scram on Low Reactor Pressure Vessel Level

No new commitments are being made in this submittal.

Should you have any questions or require additional information, please contact Mr. Ertman III Bennett III, Manager-Nuclear Licensing, at (734) 586-4273.

Peter Dietrich Senior Vice President and Chief Nuclear Officer

Enclosure: Licensee Event Repo1i No. 2022-001, Reactor Scram on Low Reactor Pressure Vessel Level

cc: NRC Project Manager NRC Resident Office Regional Administrator, Region III Enclosure to NRC-22-0013

Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43

Licensee Event Report (LER) No. 2022-001 Reactor Scram on Low Reactor Pressure Vessel Level

Abstract

At 1700, on February 4, 2022 the reactor automatically scrammed due to low reactor water level. The low reactor water level occurred as a result of a loss of feedwater while removing the south reactor feed pump (SRFP) from service. The SRFP was being removed from service per operating procedures as the plant was being shutdown to enter a refueling outage. When reducing speed on the SRFP, the north reactor feed pump (NRFP) increased speed and tripped on low suction pressure. The SRFP was unable to maintain reactor water level as the pump was in manual control at a reduced speed. The reactor water level was restored and then maintained at normal level following the scram using the condensate/feedwa ter system. Decay heat was removed through the main steam system to the main condenser. All control rods fully inserted into the reactor core. The scram was not complex. A Root Cause Evaluation is being conducted after completion of the refueling outage. (08-2 0 20) Estimated b u rden per response lo compl y with this mandatory collection request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are irKDp003ted ilto the licensing process an d fed back lo indushy. Send comments LICENSEE EVENT REPORT (LER) regarding burden estimate lo the FOIA, Library, and Information Collections Branch (T-6 A10M), U. S.

Nuclear Regulato ry Coomission, Wash ington, DC 20555--0001, or by e-mail lo CONT INUATION S H EET lnfooolleds.Resource@nrc.gov, and the 0MB reviewer at 0MB Office of Information and Regu latory Affairs, (3150-0104). Attn: Desk Officer for the Nuclear Regula tory Commission, 725 17th Street NW,

(See NUREG - 1022, R.3 for instruction and guidance for comp leting this form Washington, DC 20503; e--ma i: oira submission@omb.eop.gov. The NRG m ay not conduct or http://www.nrc.gov/read inq -rm/doc -collections/nu reqs/staff/sr1022/r3/) sponso r, and a person is no t requ ired to respond to, a col lection of i,forrnation unless the documen t requesting or requ iring the oo lection displays a currentiy valid 0MB con trol number.

1. FACI LI TY NA ME 2. DOCKET NU MBER 3. LER NU MBER Y EAR SE QU EN T IA L REV NUM BER NO.

IFerm i 2 I 05000- 1 341 ~ -I 001 1-0

NARRAT IVE

INITIAL PLANT CONDITIONS

Mode-1 Reactor Po w er - 57.9 %

There w ere no structures, systems, or co m ponents that w ere inoperab le at the start of this event that contr ibuted to this event. Ho w ever, the plant had known va lve leak-by in the Feedw ater system do wn stream of the RFPs that may have contributed to the Feedw ater system ' s response.

DESCRIPTION OF THE EVENT

At 1700 on February 4, 2022 the plant w as in Mode 1 operating at 57.9% reactor po w er. Operations personne l w ere in the process o f shutting down the plant to commence the refueling outage ( RF 21 ). Whi le low ering speed on the south reactor feed pump (SRFP ) [SJ] to remove it from service the north reactor feed pump (NRFP) tripped on lo w suction pressure [J K] and reactor w ater level decreased to Level 3, the reactor trip [JD] set po int of 173 inches above the top of active fue l (T AF ) [AC]. Reactor w ater level w as recovered initially by the SRFP. The SRFP w as later tripped manua lly due to increasing vibration levels. The Reactor Pressure Vesse l (RPV ) pressure w as lo w ered u sing the pressure regu lator to approximate ly 700 psig to maintain RPV injection with the heater feed pumps [SK]. Main Steam Iso lation Valves remained open and decay heat w as removed through the main steam system to the ma in condenser [SG].

The fo llo w ing actuation signals w ere generated w ith the reactor w ater level trip (Level 3 ), Automat ic Depressurization System (ADS ) rece ived a perm issive signal and Traversing In-co re Probes (T !Ps) rece ived a retraction signa l ho w ever TIPs did not move as they w ere a lready retracted. The fo llo w ing Primary Containment isolation signals resulted from RP V Leve l 3 signal: Group 4 RHR Shutdo w n Cooling and Head Spray (already isolated), Group 13 Drywe ll Sumps isolated and Group 15 Traversing In-Core Probe System (a lready isolated). A ll actuations and iso lations occurred as expected. All systems responded as expected. No Emergency Core Cooling Systems (ECCS ) actuated or w ere required.

Due to the reactor protection system actuat ion w hile critical, this eve n t w as reported as a non-emergency notification ( EN 55730 ) per 10 CFR 50.72 (b )( 2) (iv )( B ). The low reactor w ater level a lso caused primary containment [JM] (Groups 4, 13 and 15) isolation signa ls. The Primary Containment Group isolations not ification w as reported under 10 CFR 50.72 (b )( 3 )

(iv)( A ). Th is Licensee Even t Report (LER ) is made per 10 CFR 50.73 (a)(2 )(iv) (A ) any event or cond ition that results in actuation of the reactor protection system w hen the reactor is cr itica l and the assoc iated group isolations that occurred.

SIGN IFICANT SAFETY CONSEQUENCES AND IMPLICATIONS Reactor w ater leve l decreased to the Level 3 reactor trip set point of 173 inches above the Top of Active Fue l (T AF ) [AC].

Reactor w ater level w as restored through the Feedw ater system prior to reach ing level 2 setpoint of 11 1 inches above TAF. T he low est reactor w ater level observed dur ing the trans ient w as 117.3 inches above TAF. No Emergency Core Coo ling System [JE] actuated or w as requ ired to actuate for this scram. High Pressure Coo lant Injection [BJ], Reactor Core Iso lation Cool ing [BN] and Automat ic Depressurization System [JC] w ere ava ilable.

The Updated Fina l Safety Analysis Report ( UFSAR ) w as review ed for app licable transients sim ilar to this event and this event is bounded by existing accident ana lysis.

The reactor tripped o n lo w w ater level and the p lant responded as designed. As such, there w as no impact to the hea lth and safety of the pub lic or plant personne l.

NRG FO RM 366A (08-2020 ) Page 2 of 3 (08-2 0 20) Estimated b u rden per response lo compl y with this mandatory collection request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are irKDp003ted ilto the licensing process an d fed back lo indushy. Send comments LICENSEE EVENT REPORT (LER) regarding burden estimate lo the FOIA, Library, and Information Collections Branch (T-6 A10M), U. S.

Nuclear Regulato ry Coom ission, Wash ington, DC 20555--0001, or by e-mail lo CONT INUATION S H EET ln fooolleds.Resource@nrc.gov, an d the 0MB reviewer at 0MB Office of Information and Regu latory Affairs, (3150-0104). Attn: Desk Officer for the Nuclear Regula tory Commission, 725 17th Street NW,

(See NUREG - 1022, R.3 for instruction and guidance for comp leting this form Washington, DC 20503; e--ma i: oira submission@omb.eop.gov. The NRG m ay not conduct or http://www.nrc.gov/read inq -rm/doc -collections/nu reqs/staff/sr1022/r3/) sponso r, and a person is no t requ ired to respond to, a col lection of i,forrnation unless the documen t requesting or requiring the oo lection displays a currentiy valid 0MB con trol number.

1. FACI LI TY NA ME 2. DOCKET NU MBER 3. LER NU MBER Y EAR SE QU ENTIA L REV NUM BER NO.

IFerm i 2 I 05000- 1 341 ~ -I 001 1-0

NARRAT IVE

CAUSE OF THE EVENT

The reacto r scram w as caused by low reactor w ater level caused due to the NRFP trip on low suct ion pressure. The SRFP w as unab le to immediately recover RPV leve l due to be ing in Manua l control.

The in itial reactor scram investigation discovered some conditions that may have contr ibuted to the Feedw ater system response. Those a re the System Ope rating Procedure (SOP ) 23. 107 " Reactor Feedw ater and Condensate Systems,"

contained insufficient gu idance for impact of lo w RFP suct ion pressure w hen shutting do wn a RFP and valve leak in the Feedw a ter system do w nstream of the RFPs may have contr ibuted. A Root Cause Eva luat ion is being conducted after complet ion of the refueling outage. The Root Cause w ill determine w hether the feedw ater system design and response w ere contr ibut ing factors to the reactor trip.

CORRECT IVE ACT IONS

The leaking Feedw ater Heater Number 6 Outlet to Condenser Isolation Motor Operated va lve that may have contributed to the Feedw ate r system ' s response is be ing cut out and replaced during the refueling outage.

A Root Cause Evaluation is being conducted after completion of the refueling outage.

PREVIOUS OCCURRENCES

Th is plant shutdo w n w as compa red to two previous plant shutdo w ns performed w hen entering Forced Outage 2 1-01 and Refueling Outage 19. A review of these two prior plant shutdo w ns determined that the RFP shutdo w n occurred at lo w er reacto r po w er levels and higher RFP suction pressures. These sim ilar prior plant shutdo w ns did not resu lt in a reactor trip or the RFP trip.

NRG FO RM 366A (08-2020 ) Page 3 of 3