ML24093A286
| ML24093A286 | |
| Person / Time | |
|---|---|
| Issue date: | 04/23/2024 |
| From: | Greg Werner NRC/RGN-IV/DORS/PBB |
| To: | John Monninger Region 4 Administrator |
| References | |
| MD 8.3 | |
| Download: ML24093A286 (1) | |
Text
April 23, 2024 MEMORANDUM TO:
John D. Monninger, Regional Administrator THRU:
Geoffrey B. Miller, Director Division of Operating Reactor Safety FROM:
Gregory E. Werner, Branch Chief Projects Branch B Division of Operating Reactor Safety
SUBJECT:
MANAGEMENT DIRECTIVE 8.3 EVALUATION FOR COMANCHE PEAK NUCLEAR POWER PLANT UNIT 2 MANUAL REACTOR TRIP ON MARCH 17, 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 Comanche Peak Nuclear Power Plant Unit 2, for a manual reactor trip due to a feedwater transient. Staff performed this evaluation to determine the risk significance of the event in order to determine the appropriate level of NRC response. Based on this evaluation, staff 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:
Gregory Werner, PBB / DORS 817-200-1141 Signed by Monninger, John on 04/23/24
ML24093A286 SUNSI Review By: GEW ADAMS Yes No Non-Sensitive Sensitive Publicly Available Non-Publicly Available Keyword:
NRR-123 OFFICE SPE:DORS/B SRI:DORS/B SRA:DORS C:DORS/B D: DORS RIV:RA NAME DProulx JEllegood RDeese GWerner GMiller JMonninger SIGNATURE
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DATE 04/02/24 04/03/24 04/02/24 04/03/24 04/04/24 04/23/24 MANAGEMENT DIRECTIVE 8.3 DECISION DOCUMENTATION FORM (Deterministic and Risk Criteria Analyzed)
PLANT:
Comanche Peak Unit 2 EVENT DATE:
March 17, 2024 RESPONSIBLE BRANCH CHIEF:
Gregory Werner EVALUATION DATE:
March 21, 2024 BRIEF DESCRIPTION OF THE SIGNIFICANT OPERATIONAL EVENT OR DEGRADED CONDITION:
Comanche Peak Unit 2 was manually tripped at 3:15 pm CDT, on March 17 because of a feedwater transient. Prior to the reactor trip, the main feedwater pump B control system was in an alarm status. Subsequently, the servo lost power and main feedwater pump B tripped.
The plant automatically decreased power; however, main feedwater pump A did not increase pumping capacity quickly enough to avoid a reactor trip due to low steam generator water level. The operations crew manually tripped the reactor at 3:15 pm CDT. Following the reactor trip, all systems functioned as expected. Unit 2 stabilized in Mode 3 and reactor temperature was controlled by diverting steam to the main condenser. The trip was uncomplicated, and all safety systems responded as required.
This trip was similar to a Unit 2 automatic reactor trip that occurred on March 12, 2024, due to loss of main feedwater pump A. At the time of the loss of feedwater pump A, plant staff were changing the servo filters on the pump, causing the pump speed to decrease, and reducing feedwater flow. Main feedwater pump B responded slowly and did not increase pumping capacity as expected. The operators initiated a manual decrease in power; however, the reactor automatically tripped on low steam generator water level before the reactor operator could manually trip the reactor. Following the reactor trip, auxiliary feedwater automatically actuated, and steam was directed to the main condenser to control reactor temperature, as expected based on plant conditions. Operators stabilized the plant at normal operating temperature and pressure. The trip was uncomplicated, and all safety systems responded as required.
Y/N DETERMINISTIC CRITERIA Involved operations that exceeded, or were not included in, the design bases of the facility N
Remarks: The event was within the design basis of the facility.
Involved a major deficiency in design, construction, or operation having potential generic safety implications N
Remarks: The event did not reveal any major deficiencies, nor did it have generic safety implications.
Led to a significant loss of integrity of the fuel, primary coolant pressure boundary, or primary containment boundary of a nuclear reactor N
Remarks: There was not a loss of any barriers during this event Led to the loss of a safety function or multiple failures in systems used to mitigate an actual event N
Remarks: All safety functions performed as expected during this event and multiple failures in systems did not occur.
Involved possible adverse generic implications N
Remarks: The trip did not have generic safety implications Involved significant unexpected system interactions Y
Remarks: The main feedwater pumps did not respond as expected. The licensee had previously adjusted the automatic response of the feedwater pumps; however, those adjustments were not effective. In 2022 for Unit 1, and 2023 for Unit 2, the main feedwater control system was modified to incorporate a new Siemens control system designated as the T3000. In completing this modification, the licensee did not correctly translate the new control system design into the overall feedwater system design. In particular, the signal from the T3000 to the main feedwater pump servos was incorrectly set.
Additional troubleshooting and wiring modifications were completed. The licensee was unable to fully test the changes to ensure the main feedwater pumps responded as designed. On March 12, 2024, the control system for main feedwater pump A responded improperly, resulting in low feedwater and a subsequent reactor trip due to low steam generator level. Similarly, on March 17, the control system for main feedwater pump B responded improperly resulting in a trip of the feedwater pump followed by a manual reactor trip. This was indicative of a repeat failure of the main feedwater pump controllers to maintain steam generator levels as required.
Involved repetitive failures or events involving safety-related equipment or deficiencies in operations N
Remarks: The event did not involve failures of safety-related equipment or deficiencies in operations.
2 Y/N DETERMINISTIC CRITERIA Involved questions or concerns pertaining to licensee operational performance N
Remarks: Operators responded appropriately to the event.
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. The characteristics of the event can be addressed through NRCs normal inspection and oversight program.
Emergency Preparedness, Radiation Protection, and/or Security/Safeguards Deterministic Criteria N
Remarks: 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 BY: Rick Deese DATE: March 28, 2024 Brief description for the basis of the assessment (may include assumptions, calculations, references, peer review, or comparison with licensees results):
The analyst utilized the Comanche Peak plant-specific SPAR model, Version 8.82, run on SAPHIRE, Revision 8.2.9, to estimate the risk associated with this event. The following assumptions were made:
1.
The transient (or reactor trip) initiating event frequency was adjusted to reflect the recent operating experience of the units. A transient initiating event frequency value of 2.018/year was used in place of the nominal pressurized water reactor transient initiating event frequency of 0.518/year. The analyst noted that the site has experienced 3 reactor trips within the last calendar year, each of which initiated from main feedwater pump trips or failures. The analyst considered that performance and control of the main feedwater pumps of both units has been unreliable to the point that trips or failures of these pumps would lead to the increased frequency of a transient (reactor trip) above and beyond the baseline transient initiating event frequency level. The state of the main feedwater pumps over the past year was assumed so marked that performing a Bayesian update for this condition would not be representative. The analyst invoked a simple additive adjustment of 1.5/year to the transient initiating event frequency based on 3 feedwater anomaly-initiated reactor trips on 2 units in 1 year.
2.
Basic Event MFW-XHE-XL-TRIP, Operator Fails to Restart MFW Flow, was adjusted to a new probability of 5.0x10-1 to reflect increased difficulty in restarting the main feedwater pump system during events due to performance and control issues.
The baseline probability in the SPAR model is 4.0x10-2 and was developed using the SPAR-H human reliability analysis methodology. For the baseline value, the basic event was analyzed as action-only with barely adequate time available, high stressors, moderate complexity, and all other performance shaping factors nominal.
The analyst adjusted complexity to high and procedures to available but poor, to reflect the conditions imposed on the operators of attempting to restart main feedwater flow with challenges from the control issues stemming from the recent system performance.
The analyst quantified the SPAR model using the above assumptions to obtain an incremental conditional core damage probability of 2.0x10-6. The dominant core damage sequences included failures of the auxiliary feedwater system, feed-and-bleed equipment, and offsite power.
This evaluation includes consideration of risk from internal events only and does not include consideration of additional risk from external events, such as seismic and fire events.
4 CONDITIONAL RISK ASSESSMENT Licensee Results:
The licensee modeled the condition as a common cause failure of the main feedwater pumps along with adjusting the transient initiating event frequency to 1.0. With these assumptions, the licensee obtained an incremental conditional core damage probability of approximately 3.0x10-7. The difference in the results between the NRCs SPAR model and the licensees model are likely due to the different conditions ran and any modeling disparities between the PRA models.
THE ESTIMATED INCREMENTAL CONDITIONAL CORE DAMAGE PROBABILITY (CCDP) IS:
2.0x10-6 WHICH PLACES THE RISK IN THE RANGE OF:
No additional inspection / Special inspection overlap 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 risk being low in the overlap range and because the resident inspection staff has been evaluating various issues with the main feedwater pump system, Reactor Projects Branch B will follow up on this event using baseline inspection samples (71152A, Problem Identification and Resolution and 71111.18, Plant Modifications). The Branch will review the licensees root cause analysis of the loss of the main feedwater pump, with particular attention on the design modification that installed the T3000 control system. The inspection will include ensuring that the licensee has properly calibrated the system in order for the plant to respond to a loss of a single feedwater pump without a reactor trip and is consistent with the analysis with UFSAR.
BRANCH CHIEF REVIEW:
Gregory E. Werner D
DATE: April 3, 2024 DIVISION DIRECTOR REVIEW:
Geoffrey Miller ADAMS ACCESSION NUMBER:
EVENT NOTIFICATION REPORT NUMBER (as applicable):
E-mail to NRR_Reactive_Inspection@nrc.gov Signed by Werner, Gregory on 04/03/24 Signed by Miller, Geoffrey on 04/04/24 DATE: April 4, 2024