05000499/LER-2024-001, Automatic Reactor Trip and Actuation of Two of Three Emergency Diesel Generators

Automatic Reactor Trip and Actuation of Two of Three Emergency Diesel Generators

ML24184C083
Person / Time
Site:	South Texas
Issue date:	07/02/2024
From:	Harshaw K South Texas
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NOC-AE-24004049, STI: 35612972 LER 2024-001-00
Download: ML24184C083 (1)

LER-2024-001, Automatic Reactor Trip and Actuation of Two of Three Emergency Diesel Generators

Event date:
Report date:
4992024001R00 - NRC Website

text

July 2, 2024 NOC-AE-24004049 10 CFR 50.73 STI: 35612972 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 South Texas Project Unit 2 Docket No. STN 50-499 Licensee Event Report 2024-001-00 Automatic Reactor Trip and Actuation of Two of Three Emergency Diesel Generators

Pursuant to reporting requirements in 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(2)(v)(D),

STP Nuclear Operating Company (STPNOC) hereby submits the attached South Texas Project Unit 2 Licensee Event Report 2024-001-00 for an event or condition that: resulted in the manual or automatic actuation of any systems listed in paragraph (a)(2)(iv)(B), and could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

The event did not have an adverse effect on the health and safety of the public.

There are no commitments in this submittal.

If there are any questions regarding this submittal, please contact Chris Warren at (361) 972-7293 or me at (361) 972-4778.

Kimberly A. Harshaw Executive Vice President and Chief Nuclear Officer

Attachment: Unit 2 LER 2024-001-00, Automatic Reactor Trip and Actuation of Two of Three Emergency Diesel Generators

cc:

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 E. Lamar Boulevard Arlington, TX 76011-4511 NOC-AE-24004049 Attachment

Attachment

Unit 2 LER 2024-001-00

Automatic Reactor Trip and Actuation of Two of Three Emergency Diesel Generators

A. Reportable Event Classification

This event is reportable pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in the valid actuation of the Reactor Protection System (RPS) including reactor trip (10 CFR 50.73(a)(2)(iv)(B)(1)) and automatic actuation of Unit 2 Standby Diesel Generators (SBDG) 21 and 23 (10 CFR 50.73(a)(2)(iv)(B)(8)). This event is also reportable pursuant to 10 CFR 50.73(a)(2)(v)(D) due as an event or condition that could have prevent fulfillment of a safety function of structures or systems that are needed to mitigate the consequences of an accident.

B. Plant Operating Conditions Prior to Event

Prior to the event, Unit 2 was in Mode 1 at 15% power.

C. Status of Structures, Systems, and Components That Were Inoperable at the Start of the Event that Contributed to the Event

At the start of the event, there were no structures, systems, or components (SSCs) that were INOPERABLE that contributed to the event.

D. Background Information

Unit 2 was in Mode 1 at 15% power following completion of a refueling outage. The reactor automatically tripped due to a unit auxiliary transformer lockout. During the trip, all control rods fully inserted and all three Engineered Safety Feature (ESF) busses were energized by SBDGs 21 and 23 (Trains A and C) and Standby Transformer 2 (Train B). All equipment responded as expected except for Steam Generator Power Operated Relief Valve (PORV) 2C, which fully opened when the manual control was depressed slightly, and Load Center E2A1 supply breaker, which did not close automatically following the LOOP and automatic Engineered Safety Feature (ESF) sequencing.

Load Center E2A1 provides power to several components with an active safety function, including: Steam Generator (S/G) 2A auxiliary feedwater (AFW) flow regulator valve, hydraulic pump motor for S/G Power Operated Relief Valve (PORV) 2A, and Train 2A Diesel Generator Emergency Supply Fan 21A (Supply Fan 21A) for SBDG 21. When Load Center E2A1 breaker failed to close, the capability to provide AFW to S/G 2A was lost and S/G PORV 2A remained in the fully closed position. Supply Fan 21A provides adequate forced ventilation cooling flow for SBDG 21 to operate over its mission time.

While SBDG 21 operated successfully for the 26 minutes that Load Center E2A1 supply breaker was open, the capability for SBDG 21 to meet its design functions for its design mission time was not ensured. This impact to SBDG 21 affects S/G PORV 2D, which was powered by SBDG 21. The reliability of S/G PORV 2D to meet its design functions during the mission time could not be ensured due to the limited capability of SBDG 21 to perform its functions without forced ventilation.

Unrelated to the Load Center E2A1 power loss, PORV 2C was declared INOPERABLE due to erratic and unreliable control when placed in Manual. The PORV would not modulate to control pressure but would instead act like an on/off valve (either fully opened or closed).

The following three design basis accident conditions were evaluated against the equipment failures described above: Loss of Normal Feedwater with Loss of Offsite Power (LONF/LOAC), Steam Generator Tube Rupture (SGTR) with Loss of Offsite Power, and Small Break Loss of Coolant Accident (SBLOCA) with Loss of Offsite Power.

For a design basis LONF/LOAC, AFW trains 2B, 2C, and 2D would have been able to provide the required AFW flow rates; however, the heat removal capability of S/Gs 2C and 2D was questionable due to inoperability of PORV 2C and the inability of SBDG 21 to reliably provide power to PORV 2D as a result of Supply Fan 21A. Accordingly, only S/G 2B would have been fully capable of responding to the event.

For a design basis SGTR with Loss of Offsite Power or SBLOCA with Loss of Offsite Power, S/G 2A was not available to provide heat removal due to loss of AFW flow to the S/G. The heat removal capability of S/Gs 2C and 2D was questionable due to inoperability of PORV 2C and the inability of SBDG 21 to reliably provide power to PORV 2D. Only S/G 2B would have been fully capable of performing heat removal for rapid cooldown and mitigating the consequences of each accident.

E. Narrative Summary of the Event

Timeline (Note: All times are in Central Daylight Time)

05/12/2024 (1641) - Switchyard breakers Y590 and Y600 tripped. Unit 2 reactor tripped and EDGs 21 and 23 automatically actuated and sequenced on a Loss of Offsite Power. The following busses were deenergized: 13.8kV Auxiliary busses 2F, 2G, 2H, and 2J, 13.8kV Standby busses 2F and 2H, and Load Center 2W.

05/12/2024 (1641) - All four Reactor Coolant Pumps (RCP) lost power. Entered Technical Specification (TS) 3.4.1.2: "At least two of the reactor coolant loops listed below shall be OPERABLE and with two reactor coolant loops in operation when the reactor trip system breakers are closed and one reactor coolant loop in operation when the reactor trip system breakers are open", Action 'C': "With no reactor coolant pump in operation, suspend operations that would cause introduction into the Reactor Coolant System (RCS) of coolant with boron concentration less than required to meet Shutdown Margin of Limiting Condition of Operation (LCO) 3.1.1 and immediately initiate corrective action to return the required reactor coolant loop to operation."

05/12/2024 (1641) - Reactor Coolant Loop 'B' declared INOPERABLE due to RCP 2B #2 Seal exceeding its leak-off rate limit.

05/12/2024 (1641) - Declared loss of two offsite sources due to two ESF busses not connected to a load tap changer in Auto. Entered TS 3.8.1.1.e: "With two of the required offsite A.C. circuits inoperable, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore at least one of the inoperable offsite sources to operable status or apply the requirements of the Configuration Risk Management Program (CRMP), or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."

05/12/2024 (1641) - Load Center E2A1 supply breaker failed to close automatically following LOOP and automatic ESF sequencing. Class 1E 480V Load Center E2A1 is de-energized. Entered TS 3.8.3.1.a, Action 'A': "With one of the required trains of A.C. ESF busses not fully energized, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reenergize the train or apply the requirements of the CRMP, or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

05/12/2024 (1645) - Isolated Main Steam using safety grade switches from the Control Room.

05/12/2024 (1705) - Steam Generator PORV 2C placed in manual for venting, PORV opened to full travel without required input from operator. Steam Generator PORV 2C declared INOPERABLE.

05/12/2024 (1707) - Load Center E2A1 supply breaker automatically closed after placing it in "pull to lock" and returning the hand switch to Automatic. Class 1E 480V Load Center E2A1 is energized. Exited Technical Specification 3.8.3.1.a, Action 'A'.

05/12/2024 (1729) - Energized Standby bus 2F from Standby 2 transformer (restoration of off-site power to ESF Bus A).

05/12/2024 (1733) - Energized Standby bus 2H from Standby 1 transformer (restoration of off-site power to ESF Bus C).

05/12/2024 (1745) - Energized Auxiliary bus 2F from Standby 2 transformer.

05/12/2024 (1747) - Energized Auxiliary bus 2H from Standby 1 transformer.

05/12/2024 (1750) - Auxiliary bus 2J re-energized from off site power.

05/12/2024 (1753) - Auxiliary bus 2G re-energized from off site power.

05/12/2024 (1758) - Started RCP 2D. Reactor Coolant Loop D and its associated steam generator and RCP are in operation with the reactor trip breakers open. Exited TS 3.4.1.2, Action 'C'.

05/12/2024 (1917) - Started Reactor Coolant Pump 2A for plant stabilization

05/12/2024 (1918) - Load Center 2D1 energized from off site power

05/12/2024 (1921) - Load Center 2D2 energized from off site power

05/12/2024 (1950) - Completed NRC 4-hour notification under 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) for EDG 21 and 23 actuations. Event Notification 57124.

F. Method of Discovery

The event was self-revealing when the generator backup distance relay "21/G1", time delay relay "62G1", generator lockout relay "86G1" and the main transformer lockout relay "86SY" were flagged in the Unit 2 Control Room. The primary and backup pilot wire relays had the "DTT Key" light illuminated, indicating a transfer trip to the switchyard relays.

A. Failure Mode, Mechanism, and Effects of Failed Components

The failed components in this event were: Generator backup relay "21/G1", S/G PORV 2C, and Load Center E2A1 supply breaker. The inadvertent actuation of relay "21/G1" resulted in an automatic trip of the reactor and actuation of the RPS followed by SBDGs 21 and 23. Inoperability of S/G PORV 2C and failure of Load Center E2A1 breaker to close resulted in a condition where only S/G 2B was fully capable of providing heat removal during a LONF/LOAC, SGTR with Loss of Offsite Power, or a SBLOCA with Loss of Offsite Power. A supplement to this report will be provided detailing additional failure mechanisms upon completion of the causal analysis.

B. Cause of Component Failure

Causal analysis is ongoing. A supplement to this report will be provided detailing causes upon completion of the causal analysis.

C. Systems or Secondary Functions That Were Affected by the Failure of Components with Multiple Functions

The following Unit 2 SSCs were affected by failure of Load Center E2A1 supply breaker to close or by the failure of Supply Fan 21A for SBDG 21. There were no systems or secondary functions affected by the inoperability of PORV 2C.

- Train 'A' Emergency Core Cooling System, including Residual Heat Removal (RHR)

- Train 'A' Essential Cooling Water

- Train 'A' Essential Chilled Water

- Train 'A' Component Cooling Water

- Train 'A' Control Room Envelope HVAC

- Train 'A' Containment Spray System

- Train 'C' RHR

- S/G PORV 2A

- S/G PORV 2D

- Pressurizer Heater Group 2A

- Electrical Auxiliary Building HVAC

- Auxiliary Feedwater Pump 21

- Reactor Containment Fan Cooler

D. Failed Component Information

System: Main Steam System {SB}

Component: Valve, Control, Pressure {PCV}

Manufacturer: {Control Components Inc}

Model: {35199-1-2}

System: Main Generator Output Power System {EL}

Component: Backup Distance Relay {RLY-21}

Manufacturer: Westinghouse {W120}

Model: {719B196A11}

System: Low-Voltage Power System {ED}

Component: Breaker {BKR}

Manufacturer: Westinghouse {W120}

Model: {DS-416}

System: Low-Voltage Power System {ED}

Component: Hand switch {HS}

Manufacturer: General Electric {G080}

Model: {KA-2176-SD30}

III. Analysis of Event

A. Safety System Responses that Occurred

The Reactor Protection System actuated, automatically tripping the reactor following the inadvertent relay actuation.

SBDGs 21 and 23 automatically started within the required time frame.

B. Duration of the Safety System Inoperability

Load Center E2A1 supply breaker failed to automatically close at 1641 on May 12, 2024. The breaker automatically closed at 1707. This was a total of 26 minutes. PORV 2C was INOPERABLE from 1705 on May 12, 2024, and was declared OPERABLE at 1031 on May 16, 2024. This was a total of 3 days, 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />, and 26 minutes.

C. Safety Consequences and Implications

The Incremental Core Damage Probability (ICDP) and Incremental Large Early Release Probability (ILERP) of the event for the 66 minutes where Load Center E2A1 and SG PORV 2C are concurrently inoperable is 2.28E-07 and 1.19E-09, respectively. The exposure time of 66 minutes is the time of the reactor trip to the time that the 13.8 kV buses 2F and 2H are energized from the Standby Transformer 2. The ICDP and ILERP are below the threshold of 1E-06 and 1E-07 and is therefore considered of very low safety significance.

IV. Cause of Event

Causal analysis is ongoing. A supplement to this report will be provided detailing causes upon completion of the causal analysis.

Completed Corrective Actions

1. Relay "21/G1" replaced with a spare.

2. Installation of the "62-G1" time delay relay and "21/G1" relay immediately before the field breaker is closed.

3. Repairing the CP010 door to minimize vibrations from the door closure.

Causal analysis is ongoing. A supplement to this report will be provided detailing additional corrective actions upon completion of the causal analysis.

VI. Previous Similar Events

The following previous similar events were identified for S/G PORVs:

1. Condition Report (CR) 24-4183: SG 2C PORV did not respond to the Close (down) pushbutton being depressed for approximately 11 seconds during 0PSP05-MS-7411L. SG 2C PORV did travel full closed after the delay.

2. CRs 23-10139 and 23-10508: STP received a 10 CFR Part 21 notification (Certrec Event # 56821) from Enertech:

Notification of potential defect for Steam Generator PORV potentiometer part number D2060S. A number of returned potentiometers to Enertech exhibited inconsistent resistance values at certain stroke positions. The potentiometers are used in modulating actuators. Per Enertech STP is impacted. A plant impact form was completed under CR 23-10139.

This CR is tracking additional actions needed to be taken to correct the vulnerabiliites with the impacted PORV potentiometers for Steam Generators 1C, 2A, 2B, and 2C.

3. CR 22-12288: Received the following silent Plant Computer alarms: MSZE7421 SG 2B PORV POSIT at 2.1% and slowly increasing (~1% / 30 mins) and MSZC7421 SG 2B PORV OPEN. No observable increases in Steam Flow from S/G 2B or effects on the Reactor Coolant System were observed.

4. CR 21-10382: While venting S/G 1D during performance procedure 0POP03-ZG-0007, "Plant Cooldown," while opening S/G 1D PORV and with the Valve Position Indicator (VPI) indicating approximately 90% open, we observed VPI peg low to 0% and then return to 90% two times. No other indications occurred to indicate the valve moved.

5. CR 21-8976: During performance of WAN# 605160, Operations used the Operator Interface Module (OIM) up pushbutton to open 2C PORV. After holding the up pushbutton for an adequate time for a full open stroke, the valve never moved.

6. CR 18-5534: SG 'C' PORV indication on CP-006, plant computer point MSZE7431 showing valve position is oscillating from as high as 99% to as low at 9% open. Investigate cause of valve position indication oscillation. Operator reports no stem movement locally and indicating full open.

No previous similar events were identified for Load Center E2A1 supply breaker 2E or relay "21/G1".