05000341/LER-2024-001, Automatic RPS Scram on High RPV Pressure While Attempting to Lower Generator Output During RF-22
| ML24136A211 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/15/2024 |
| From: | Peter Dietrich DTE Electric Company |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NRC-24-0025 LER 2024-001-0 | |
| Download: ML24136A211 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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) 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 |
| 3412024001R00 - NRC Website | |
text
Peter Dietrich Senior Vice President and Chief Nuclear Officer DTE Electric Company 6400 N. Dixie Flighway, Newport, MI48166 Tel: 734.586.6515 Email: peter.dietrichidteenergy.com DTE May 15, 2024 10 CFR 50.73 NRC-24-0025 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 Report (LER) No. 2024-001 Pursuant to 10CFR50.73(a)(2)(iv)(A), DTE Electric Company (DTE) is submitting LER No.
2024-001, Automatic RPS SCRAM on High RPV Pressure While Attempting to Lower Generator Output During RF-22.
No new commitments are being made in this submittal.
Should you have any questions or require additional information, please contact Mr. Eric Frank, Manager - Nuclear Licensing, at (734) 586-4772.
Sincerely, Peter Dietrich Senior Vice President and Chief Nuclear Officer
Enclosure:
Licensee Event Report No. 2024-001, Automatic RPS SCRAM on High RPV Pressure While Attempting to Lower Generator Output During RF-22 cc:
NRC Project Manager NRC Resident Office Regional Administrator, Region III
Enclosure to NRC-24-0025 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Licensee Event Report (LER) No. 2024-001 Automatic RPS SCRAM on High RPV Pressure While Attempting to Lower Generator Output During RF-22
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY0OMB: NO. 3150-0104 EXPIRES: 0413012027 (0-222)Estimated burden per response to complyowith sis mandatory collection request:80s ours.
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- 3. Page Fermi 2 Ej 052 00341 1 OF 3
- 4. Title Automatic RPS SCRAM on High RPV Pressure While Attempting to Lower Generator Output During RF-22
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved Mnh Dy Ya Yer Sequential Revision Month Day Year Facility Name Docket Number 1
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INITIAL PLANT CONDITIONS
Mode -1 Reactor Power - 023 There were no structures, systems, or components (SSCs) that were inoperable at the start of this event that contributed to this event.
DESCRIPTION OF THE EVENT At 0004 Eastern Daylight Time (EDT) on March 23, 2024, with the unit in Mode 1 at 23-percent power, the reactor automatically scrammed due to high Reactor Pressure Vessel (RPV) pressure when the turbine bypass valves [PCV] (which were being opened per Fermi procedure 22.000.04 "Plant Shutdown from 25-percent Power") unexpectedly closed while attempting to lower Generator [GEN] output to 55 MWe to support shutdown for a refueling outage. The SCRAM was not complex, with systems responding normally post-scram, with the exception of the pressure control system. The transient occurred while lowering Turbine Speed/Load Demand which caused arise in pressure/power until the Reactor Pressure System (RPS) [JD] Set point for Reactor Pressure High was exceeded and resulted in an Automatic Reactor Scram.
Reactor water level was maintained at normal level following expected post-scram feedwater level response. Decay heat was removed by the Main Steam System [SB] to the main condenser using manual operation of turbine bypass valves. All Control Rods were inserted into the core.
DESCRIPTION OF THE SYSTEM The Main Turbine Bypass System [JI] and Moisture Separator Reheater (MSR) [RHTR] are designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cool down. It allows excess steam flow from the reactor to the condenser without going through the turbine. The bypass capacity of the Main Turbine Bypass System and Moisture Separator Reheater is 30-percent of the Nuclear Steam Supply System [SB] rated steam flow. Sudden load reductions within the capacity of the steam bypass can be accommodated without reactor SCRAM. The Main Turbine Bypass System consists of two valves connected to a 52-inch manifold, which is between the main steam isolation valves and the turbine stop valves. Each of these valves is operated by a separate hydraulic unitized actuator. The bypass valves are controlled by the Main Turbine Pressure Regulator Control System [JI],
as discussed in the Updated Final Safety Analysis Report (UFSAR), Section 7.7.1.4. The bypass valves are normally closed, and the pressure regulator controls the turbine control valves that direct all steam flow to the turbine. If the turbine speed controls or the load limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves. When the bypass valves open, the steam flows from the 52-inch manifold through connecting piping to the condenser.
The reheating steam flow path to the moisture separator reheater provides an additional steam volume that mitigates a rapid pressure increase (e.g., from a generator load reduction).
SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
Due to the reactor protection system actuation while critical, this event was reported as a four-hour, non-emergency notification (EN 57046) per 10 CFR 50.72(b)(2)(iv)(B). Additionally, we received expected isolations for Level 3: Group 13 Drywell Sumps, Group 15 Traversing In-core Probes (TIPs) (which was already isolated) and Group 4 Residual Heat Removal - Shutdown Cooling (RHR-SDC) (which was already isolated). The Primary Containment Isolation Event is being reported under 10 CFR 50.72(b)(3)(iv)(A). Also, due to the main turbine bypass valves unexpectedly closing, this was initially reported under 10 CFR 50.72(b)(3)(v)(D), but later revised to remove this reporting criteria because this was not an event that would have prevented the fulfillment of a safety function.
Unexpected bypass valve closure during operation of turbine controls speed/load demand resulted in reactor pressure increase. This created a positive feedback loop where the pressure rise raises reactor power which further raises pressure.
The subsequent turbine control valve closure over a one-minute period accelerated the reactor pressure increase to above the reactor pressure SCRAM set point of 1093 psig and an automatic Reactor SCRAM occurred. The maximum reactor pressure reached was 1102 psig. The plant response was assessed against the descriptions provided in UFSAR Section 15.2 "Increase in Reactor Pressure", as symptoms of event are similar. While UFSAR analyses typically assume 100-percent power operation, the response of the plant to increasing pressure at the power level of this event [22.5-percent Core Thermal Power (CTP)] is consistent with Fermi's design features. Reactor Pressure - High RPS trip functioned at the correct set point and plant was recovered using feedwater and the Main Condenser as a heat sink.
As such, there was no impact to the health and safety of the public.
CAUSE OF THE EVENT
The cause of the SCRAM was the failure of the RL16 Block Load Relay which resulted in a trip of a bypass valve logic and prevented the opening of the bypass valves, in automatic, when required. The RL16 relay provides a signal to add 55 MWe (load) to the Narrow Range Speed Governor (NRSG) signal when the unit is connected to the grid. NRSG is one of 5 modes of operation of the turbine control system that maintains the speed of the turbine at 1800 RPM or the load of the main generator at a given setpoint. The remaining 4 modes are Pressure, CVOL (Turbine Flow Limit), Run Up, and Turbine Trip. Of these 5 control signals, the lowest signal, which corresponds to the lowest valve demand, is selected to be in control by a low value gate. The output of the low value gate is sent to control the steam inlet valves to the turbine. The bypass valves are positioned by the difference in the output of the low value gate and the pressure signal to maintain pressure at the 52-inch manifold. In normal operation, the pressure control signal is maintained as the low value to the low value gate and is in control.
CORRECTIVE ACTIONS
Fermi 2 has replaced the RL16 Block Load Relay.
Fermi Procedure 22.000.04 ("Plant Shutdown from 25-Percent Power") Section 6 will be revised to eliminate operating on Narrow Range Speed Governor (NRSG) during reactor shutdown due to the potential latent failure mechanism.
PREVIOUS OCCURRENCES
There are no previous occurrences.Page 3
of 3