ML24088A267
| ML24088A267 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/08/2024 |
| From: | John Dixon NRC/RGN-IV/DORS/PBD |
| To: | John Monninger NRC Region 4 |
| References | |
| EN 57042 | |
| Download: ML24088A267 (7) | |
Text
MEMORANDUM TO: John Monninger, Regional Administrator THRU: Geoffery B. Miller, Director Division of Operating Reactor Safety FROM: John L. Dixon, Branch Chief Project Branch D
SUBJECT:
MANAGEMENT DIRECTIVE 8.3 EVALUATION FOR WATERFORD STEAM ELECTRIC STATION, UNIT 3, MANUAL TRIP ON MARCH 16, 2024 Pursuant to Regional Office Policy Guide 0801, Management Directive 8.3 and Inspection Manual Chapter 0309 Reactive Team Inspection Decisions, Implementation, and Documentation for Power Reactors, the enclosed table provides the Management Directive 8.3 evaluation for determining that no additional inspection will be conducted at Waterford Steam Electric Station, Unit 3, for a manual reactor trip due to main feedwater valve and main steam isolation valve going closed for steam generator B. This event was a direct result of a failed Engineered Safety Feature Actuation System (ESFAS) relay. This was the second ESFAS relay related reactor trip within 2 years. The branch will use baseline inspection procedures for follow up inspection of this event.
Concur with Recommendation:
John D. Monninger Date Regional Administrator
Enclosures:
MD 8.3 Decision Documentation Form (Deterministic and Risk Criteria Analyzed)
CONTACT: John Dixon, DORS/PBD 817-200-1574April 8, 2024 Signed by Miller, Geoffrey on 03/29/24 Signed by Dixon, John on 04/01/24 Signed by Monninger, John on 04/08/24
J. Monninger 2
MANAGEMENT DIRECTIVE 8.3 EVALUATION FOR WATERFORD STEAM ELECTRIC STATION, UNIT 3, MANUAL TRIP ON MARCH 16, 2024 - DATED APRIL 08, 2024
DISTRIBUTION:
JMonninger, ORA JLara, ORA GMiller, DORS MHay, DORS DCylkowski, RC FGaskins, RIV/OEDO VDricks, ORA LWilkins, OCA JDrake, NRR AMoreno, RIV/OCA RAlexander, RSLO JDixon, DORS ASanchez, DORS DChilds, DORS LReyna, DORS RDeese, DORS R4-DORS-IPAT R4Enforcement NRR Reactive Inspection NRR_Reactive_Inspection@nrc.gov
DOCUMENT NAME: MANAGEMENT DIRECTIVE 8.3 EVALUATION FOR WATERFORD STEAM ELECTRIC STATION, UNIT 3 MANUAL TRIP ON MARCH 16, 2024
SUNSI Review ADAMS: Non-Publicly Available Non-Sensitive Keyword:
By: ASanchez Yes No Publicly Available Sensitive
ADAMS Accession Number: ML24088A267 OFFICE DORS:SPE DORS:SRA DORS: PDB DORS:DD RA NAME ASanchez RDeese JDixon GMiller JMonninger DATE 03/28/2024 03/29/2024 03/29/2024 03/29/2024 04/08/2024 OFFICIAL RECORD COPY MANAGEMENT DIRECTIVE 8.3 DECISION DOCUMENTATION FORM (Deterministic and Risk Criteria Analyzed)
PLANT: Waterford 3 EVENT DATE: 3-16-2024 RESPONSIBLE John Dixon EVALUATION 3-21-2024 BRANCH CHIEF: DATE:
BRIEF DESCRIPTION OF THE SIGNIFICANT OPERATIONAL EVENT OR DEGRADED CONDITION:
On March 16, 2024, at 2:49 pm CDT, Waterford Steam Electric Station, Unit 3, was operating at 100 percent power when a manual reactor trip was initiated due to main feed isolation valve No. 2, FW-184B, and main steam isolation valve No. 2, MS-124B, going closed unexpectedly. Main feedwater pump A tripped on low flow, but pump B continued to operate and supply water to steam generator No. 1. Emergency feedwater was automatically actuated and supplied water to steam generator No. 2. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip. The unit is currently stable in Mode 3. All control rods fully inserted as expected and all other plant equipment functioned as expected.
Operators completed their standard post trip actions and exited to general operating procedures. No emergency declarations were made, and key safety systems responded as expected. The resident inspector responded to the station to follow-up on the stations actions in response to the reactor trip and understand the cause(s) of the trip.
The direct cause of the main feed isolation valve No. 2 and main steam isolation valve No. 2 closing was a failure in an engineering safety feature actuation system (ESFAS) relay that provides for a main steam and main feed isolation.
In June 2022, a similar event occurred with the failure of the same type of ESFAS relay. In that event, an automatic reactor trip occurred. Following this June 2022 event, all four of these relays were replaced due to age. The licensee is evaluating this current event for Part 21 requirements, the cause of the early failure of the relay, or an anomaly in the ESFAS circuit.
Enclosure Y/N DETERMINISTIC CRITERIA
Involved operations that exceeded, or were not included in, the design N bases of the facility Remarks: Reactor trips, loss of main feedwater, and main steam isolation are part of the design basis of the facility.
Involved a major deficiency in design, construction, or operation having potential generic safety implications N Remarks: This was not a deficiency of design, construction, or operation and does not currently have generic safety implications. The relay that caused the event was replaced less than two years ago due to age.
Led to a significant loss of integrity of the fuel, primary coolant pressure boundary, or primary containment boundary of a nuclear reactor N Remarks: No loss of integrity of fuel, reactor coolant pressure boundary, or containment boundary.
Led to the loss of a safety function or multiple failures in systems used to N mitigate an actual event Remarks: No loss of safety function or multiple failures in systems.
Involved possible adverse generic implications N Remarks: The licensee is evaluating for Part 21 requirements and the cause of the premature failure as the other three relays tested satisfactorily.
Involved significant unexpected system interactions N Remarks: All systems responded as designed.
Involved repetitive failures or events involving safety-related equipment or deficiencies in operations Remarks: In June of 2022, Waterford experienced a similar event (failed ESFAS Relay, main feed and steam valve isolation on steam generator number 2 and Y automatic reactor trip). The corrective action was to replace all four relays because of their age and send the relay off for evaluation and to implement a modification that would have prevented the failure of one relay that led to the reactor trip. That modification was scheduled for the fall 2023 but was postponed.
The licensee is investigating why the current ESFAS relay failed.
N Involved questions or concerns pertaining to licensee operational performance
2 Y/N DETERMINISTIC CRITERIA
Remarks: Operator actions were in accordance with station procedures and from the prior event in 2022 they replaced all four relays and reduced the replacement frequency.
Involved circumstances sufficiently complex, unique, or not well enough understood, or involved safeguards concerns, or involved characteristics the investigation of which would best serve the needs and interests of the Commission N Remarks: This event is not complex or unique and it is understood how it occurred. There are no safeguards concerns, nor do the characteristics of the investigation best serve the needs and interests of the Commission.
Further, none of the Emergency Preparedness, Radiation Protection, and/or Security/Safeguards Deterministic Criteria could be answered in the affirmative for this event.
3 CONDITIONAL RISK ASSESSMENT
IF IT IS DETERMINED THAT A RISK ANALYSIS IS NOT REQUIRED - ENTER NA BELOW AND CONTINUE TO THE DECISION BASIS BLOCK
RISK ANALYSIS Rick Deese DATE: March 20, 2024 BY:
Brief description for the basis of the assessment (may include assumptions, calculations, references, peer review, or comparison with licensees results):
The analyst utilized the plant-specific SPAR model, Version 8.81, run on SAPHIRE, Version 8.2.9, to estimate the risk associated with this event. The following assumptions were made:
- 1. Main feed isolation valve No. 2, FW-184B closed during the initiation of the event.
This condition was modeled by setting the Basic Event MFW-HOV-OC-184B, Failure of MFW Inlet Hydraulic Valve FW-184B, to TRUE in the Events and Conditions Assessment (ECA) Workspace.
- 2. In this event, the reactor trip involved a loss of main feedwater pump A when it tripped due to low flow. This condition was modeled by setting the Basic Event MFW-TDP-FR-A, Failure of MFW A to Run, to TRUE.
- 3. Steam generator A remained available for heat removal after the event by feeding with the remaining main feed pump and discharging steam flow to the turbine bypass valves to the condenser. Steam generator B remained available for heat removal after the event by feeding with the emergency feedwater system and discharging steam through the steam generator B atmospheric relief valves as needed. The turbine bypass valves in steam header B were assumed available because the main steam system is cross connected and therefore the bypass valves remained available.
- 4. The analyst assumed no other mitigating equipment was out of service at the time and applied the zero test and maintenance change set.
The analyst quantified the SPAR model using the above assumptions to obtain a conditional core damage probability of 6.8 x 10-7. The dominant core damage sequences included reactor trips during which the safety relief valves were actuated and subsequently failed to reseat.
Licensee Results:
The analyst did not solicit the licensee for their results, but instead calculated that the licensees internal events model would estimate the conditional core damage probability to be approximately 2.5 x 10-7 by using information from the licensees internal events PRA summary document.
THE ESTIMATED CONDITIONAL CORE DAMAGE 6.8 x 10-7 PROBABILITY (CCDP) IS:
WHICH PLACES THE RISK IN THE RANGE OF: No additional inspection
4 RESPONSE DECISION
USING THE ABOVE INFORMATION AND OTHER KEY ELEMENTS OF CONSIDERATION AS APPROPRIATE, DOCUMENT THE RESPONSE DECISION TO THE EVENT OR CONDITION, AND THE BASIS FOR THAT DECISION DECISION AND DETAILS OF THE BASIS FOR THE DECISION:
Based on the low risk evaluation and no concerns about operator performance or plant response the branch recommends baseline inspection for follow up of this event.
The branch will utilize baseline inspection procedures to evaluate the licensees causal evaluation and determination of the failure of the ESFAS relay, as well as understanding the licensees strategy to remedy the known single point vulnerability.
BRANCH CHIEF John Dixon DATE: 03/21/2024 REVIEW:
DIVISION DIRECTOR Geoffrey Miller DATE: 03/29/2024 REVIEW:
ADAMS ACCESSION NUMBER:
EVENT NOTIFICATION REPORT NUMBER (as applicable):
E-mail to NRR_Reactive_Inspection@nrc.gov
5