05000341/LER-2022-001, 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) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3412022001R00 - NRC Website | |
text
April 4, 2022 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 Pdcr Didrich Senior Vice President and Chief Nucle:u Officer DTE Electric Company 6400 N. Dixie Highway, Newport, Ml 48166 Tel: 734.586.41 53 Email: petcr.dictrich@dtccnergy.co111 DTE 10 CFR 50.73 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/feedwater 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.
N/A l~~
~
t
?
/
1
INITIAL PLANT CONDITIONS
Mode-1 Reactor Power - 57.9%
There were no structures, systems, or components that were inoperable at the start of this event that contributed to this event. However, the plant had known valve leak-by in the Feedwater system downstream of the RFPs that may have contributed to the Feedwater system's response.
DESCRIPTION OF THE EVENT At 1700 on February 4, 2022 the plant was in Mode 1 operating at 57.9% reactor power. Operations personnel were in the process of shutting down the plant to commence the refueling outage (RF 21 ). While lowering speed on the south reactor feed pump (SRFP) [SJ] to remove it from service the north reactor feed pump (NRFP) tripped on low suction pressure [JK] and reactor water level decreased to Level 3, the reactor trip [JD] set point of 173 inches above the top of active fuel (T AF) [AC]. Reactor water level was recovered initially by the SRFP. The SRFP was later tripped manually due to increasing vibration levels. The Reactor Pressure Vessel (RPV) pressure was lowered using the pressure regulator to approximately 700 psig to maintain RPV injection with the heater feed pumps [SK]. Main Steam Isolation Valves remained open and decay heat was removed through the main steam system to the main condenser [SG].
The following actuation signals were generated with the reactor water level trip (Level 3), Automatic Depressurization System (ADS) received a permissive signal and Traversing In-core Probes (T!Ps) received a retraction signal however TIPs did not move as they were already retracted. The following Primary Containment isolation signals resulted from RPV Level 3 signal: Group 4 RHR Shutdown Cooling and Head Spray (already isolated), Group 13 Drywell Sumps isolated and Group 15 Traversing In-Core Probe System (already isolated). All actuations and isolations occurred as expected. All systems responded as expected. No Emergency Core Cooling Systems (ECCS) actuated or were required.
Due to the reactor protection system actuation while critical, this event was reported as a non-emergency notification (EN 55730) per 10 CFR 50.72(b)(2)(iv)(B). The low reactor water level also caused primary containment [JM] (Groups 4, 13 and 15) isolation signals. The Primary Containment Group isolations notification was reported under 10 CFR 50.72(b)(3)
(iv)(A). This Licensee Event Report (LER) is made per 10 CFR 50.73(a)(2)(iv)(A) any event or condition that results in actuation of the reactor protection system when the reactor is critical and the associated group isolations that occurred.
SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
Reactor water level decreased to the Level 3 reactor trip set point of 173 inches above the Top of Active Fuel (T AF) [AC].
Reactor water level was restored through the Feedwater system prior to reaching level 2 setpoint of 11 1 inches above TAF. The lowest reactor water level observed during the transient was 117.3 inches above TAF. No Emergency Core Cooling System [JE] actuated or was required to actuate for this scram. High Pressure Coolant Injection [BJ], Reactor Core Isolation Cooling [BN] and Automatic Depressurization System [JC] were available.
The Updated Final Safety Analysis Report (UFSAR) was reviewed for applicable transients similar to this event and this event is bounded by existing accident analysis.
The reactor tripped on low water level and the plant responded as designed. As such, there was no impact to the health and safety of the public or plant personnel.
CAUSE OF THE EVENT
The reactor scram was caused by low reactor water level caused due to the NRFP trip on low suction pressure. The SRFP was unable to immediately recover RPV level due to being in Manual control.
The initial reactor scram investigation discovered some conditions that may have contributed to the Feedwater system response. Those are the System Operating Procedure (SOP) 23.107 "Reactor Feedwater and Condensate Systems,"
contained insufficient guidance for impact of low RFP suction pressure when shutting down a RFP and valve leak in the Feedwater system downstream of the RFPs may have contributed. A Root Cause Evaluation is being conducted after completion of the refueling outage. The Root Cause will determine whether the feedwater system design and response were contributing factors to the reactor trip.
CORRECTIVE ACTIONS
The leaking Feedwater Heater Number 6 Outlet to Condenser Isolation Motor Operated valve that may have contributed to the Feedwater system's response is being cut out and replaced during the refueling outage.
A Root Cause Evaluation is being conducted after completion of the refueling outage.
PREVIOUS OCCURRENCES
This plant shutdown was compared to two previous plant shutdowns performed when entering Forced Outage 21-01 and Refueling Outage 19. A review of these two prior plant shutdowns determined that the RFP shutdown occurred at lower reactor power levels and higher RFP suction pressures. These similar prior plant shutdowns did not result in a reactor trip or the RFP trip. Page 3
of 3