IR 05000369/2025004
| ML26028A133 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  (NPF-009, NPF-017) |
| Issue date: | 02/03/2026 |
| From: | Robert Williams Division of Operating Reactors |
| To: | Pigott E Duke Energy Carolinas |
| References | |
| IR 2025004 | |
| Download: ML26028A133 (0) | |
Text
SUBJECT:
MCGUIRE NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000369/2025004 AND 05000370/2025004
Dear Edward Pigott:
On December 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at McGuire Nuclear Station. On January 14, 2026, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at McGuire Nuclear Station.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at McGuire Nuclear Station.
February 3, 2026 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Robert E. Williams, Jr., Chief Projects Branch 1 Division of Operating Reactor Safety Docket Nos. 05000369 and 05000370 License Nos. NPF-9 and NPF-17
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000369 and 05000370
License Numbers:
Report Numbers:
05000369/2025004 and 05000370/2025004
Enterprise Identifier:
I-2025-004-0023
Licensee:
Duke Energy Carolinas, LLC
Facility:
McGuire Nuclear Station
Location:
Huntersville, North Carolina
Inspection Dates:
October 1, 2025, to December 31, 2025
Inspectors:
C. Safouri, Senior Resident Inspector
F. Young, Resident Inspector
M. Kennard, Senior Operations Engineer
Approved By:
Robert E. Williams, Jr., Chief
Projects Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at McGuire Nuclear Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Prescribe Instructions Appropriate to the Circumstance for Charging Activities while in Off-normal Alignment Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000370/2025004-01 Open/Closed
[H.11] -
Challenge the Unknown 71111.15 A self-revealed Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to provide instructions or procedures for charging activities while in an off-normal chemical and volume control system (CVCS)alignment.
Failure to Correct Leakage of a Boric Acid Relief Valve Results in a Manual Reactor Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000370/2025004-02 Open/Closed None (NPP)71152A A self-revealed Green finding and NCV of 10 CFR Part 50, Appendix B, Criterion XVI,
"Corrective Action," was identified when the licensee failed to correct a condition adverse to quality (CAQ) associated with the Unit 2 boric acid blender outlet relief valve, 2NV-483, which resulted in a four gallon per minute (gpm) reactor coolant (NC) system leak outside of containment and a manual reactor trip.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000370/2025-001-00 LER 2025-001-00 for McGuire Nuclear Station,
Unit 2, Condition Prohibited by Technical Specifications Due to Emergency Diesel Ventilation Damper Malfunction 71153 Closed
LER 05000370/2025-001-01 LER 2025-001-01 for McGuire Nuclear Station,
Unit 2, Condition Prohibited by Technical Specifications Due to Emergency Diesel Ventilation Damper Malfunction 71153 Closed LER 05000370/2025-002-00 LER 2025-002-00 for McGuire Nuclear Station,
Unit 2, Reactor Trip,
Auxiliary Feedwater Actuation, and Refueling Water Storage Tank Inoperability Due to Relief Valve Leak 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at rated thermal power (RTP). On November 10, 2025, the unit experienced a secondary system transient (loss of 1C1 and 1C2 heater drain tank pumps)resulting in the unit being rapidly down powered to 85 percent RTP. Following repairs, the unit was returned to RTP on November 13, 2025. On December 11, 2025, the unit was down powered to approximately 26 percent RTP to perform a planned repair of the Unit 1 auxiliary transformer (1ATB) bushings. Following repairs, the unit was returned to RTP on December 15, 2025. On December 29, 2025, the unit experienced an automatic reactor trip due to a turbine trip (generator lockout). The unit remained in hot standby (Mode 3) to perform repairs through the remainder of the inspection period.
Unit 2 began the inspection period at RTP. On October 8, 2025, the unit was down powered to 30 percent RTP to perform repairs on the Unit 2 auxiliary transformer (2ATB) bushing and leakby from a main feedwater high pressure heater tube side relief valve. On October 9, 2025, the unit was down powered to 20 percent RTP to restore the high pressure heater to service.
On October 10, 2025, the unit was rapidly down powered and placed in hot standby (Mode 3)due to an active leak from the Unit 2 boric acid blender outlet relief valve. Following repairs, the unit was returned to RTP on October 19, 2025, and remained at or near RTP for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2 offsite power paths of the main power distribution system, on October 7, 2025
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Unit 2 auxiliary feedwater pump room, on October 3, 2025
- (2) Unit 1 train B 4160V essential switchgear room, on October 29, 2025
- (3) Unit 1 train A 4160V essential switchgear room, on November 5, 2025
- (4) Main fire pumps and main intake structure, on November 11, 2025
- (5) Unit 1 and Unit 2 diesel generator rooms, on November 21, 2025
- (6) Unit 1 auxiliary feedwater pump room, on December 6, 2025
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03) (1 Sample)
The licensee completed the annual requalification operating examinations and biennial written examinations required to be administered to all licensed operators in accordance with 10 CFR 55.59(a)(2), "Requalification Requirements," of the NRC's "Operator's Licenses." The inspector performed an in-office review of the overall pass/fail results of the individual operating examinations, the crew simulator operating examinations, and the biennial written examinations in accordance with IP 71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance." These results were compared to the thresholds established in Section 3.03, "Requalification Examination Results," of IP 71111.11.
- (1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam administered on August 21, 2025, and the biennial written examination completed on June 24, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Nuclear condition report (NCR) 02544243, Unit 2 train A diesel generator ventilation fan #1 (2A1) reverse rotation and 2A1 outside air backdraft damper failure, on October 1, 2025
- (2) NCR 02570698, Unit 2 auxiliary transformer, 2ATB, Y-phase bushing found with oil level above sight glass, on October 1, 2025
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activity to ensure configuration changes and appropriate work controls were addressed:
- (1) NCR 02570698, emergent work on Unit 2 train B auxiliary transformer, 2ATB, bushing seal leak, on October 8, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) NCR 02570532, Unit 1 train B diesel generator control air solenoid valve, 1VGSV-5170, ground, on October 1, 2025
- (2) NCR 02571813, licensee actions to maintain refueling water storage tank operability during plant cooldown for forced outage, on October 10, 2025
- (3) NCR 02571935, Unit 2 pressurizer maximum cooldown and heatup rates exceeded during cold shutdown, on October 12, 2025
- (4) NCR 02571910, Unit 2 pressurizer power operated relief valve, 2NC-33A, breaker chatter while cycling the valve, on October 21, 2025
- (5) NCR 02573295, Unit 2 pressurizer power operated relief valve drain, 2NC-270, open indication not illuminating when cycled, on November 12, 2025
- (6) NCR 02568990, evaluation of continuous volume control tank purge operation, on December 1, 2025
- (7) NCR 02568504, Unit 1 control room door pins not operating, on December 2, 2025
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated Unit 2 forced outage, M2F30A, activities from October 10, 2025, to October 17, 2025.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (4 Samples)
- (1) Electrical resistance testing of Unit 2 auxiliary transformer, 2ATB, following replacement of transformer bushing, on October 10, 2025
- (2) Bench testing and post-installation functional testing of Unit 2 boric acid blender outlet relief valve, 2NV-483, following emergent component replacement, on October 13, 2025
- (3) Unit 1 and Unit 2 train B shared control room ventilation chiller functional testing following oil pump replacement, on November 7, 2025
- (4) Unit 1 train A centrifugal charging pump functional testing following a planned preventative maintenance window, on November 11, 2025
Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) OP/2/A/6100/SU-9, "Unit 2 Mode 4 Checklist," on October 14, 2025
- (2) PT/1/A/4350/004, "Unit 1 4kV Loss of Voltage Trip Actuating Device Operational Test," on December 4, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) PT/2/A/4208/001A, "Unit 2 Train A Containment Spray Pump Performance Test," and PT/2/A/4208/002A, "Unit 2 Train A Valve Stroke Timing - Quarterly," on November 5, 2025
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
The inspectors evaluated:
- (1) An emergency preparedness drill that consisted of a steam generator tube rupture, stuck open steam generator safety valve, and fuel failure, on October 23, 2025
OTHER ACTIVITIES - BASELINE
71152A - Annual Follow-up Problem Identification and Resolution
Annual Follow-up of Selected Issues (Section 03.03) (1 Sample)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) NCR 02571346, Unit 2 boric acid blender outlet relief, 2NV-483, leakage resulting in manual reactor trip, on November 20, 2025
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02) (2 Samples)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000370/2025-001-00 and corresponding revision, LER 05000370/
2025-001-01, for McGuire Nuclear Station, Unit 2, Condition Prohibited by Technical Specifications Due to Emergency Diesel Ventilation Damper Malfunction (ADAMS Accession Nos. ML25307A321 and ML25338A159). The inspection conclusions associated with these LERs are documented in this report under Inspection Results Section 71153. These LERs are Closed.
- (2) LER 05000370/2025-002-00 for McGuire Nuclear Station, Unit 2, Reactor Trip, Auxiliary Feedwater Actuation, and Refueling Water Storage Tank Inoperability Due to Relief Valve Leak (ADAMS Accession No. ML25343A343). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71152A. This LER is Closed.
Personnel Performance (IP Section 03.03) (2 Samples)
- (1) The inspectors evaluated the licensee's emergent response to the four gpm leak from the Unit 2 boric acid blender outlet relief valve, 2NV-483, and the licensees performance, on October 10, 2025.
- (2) The inspectors evaluated the Unit 1 automatic reactor trip due to a turbine trip (generator lockout) and the licensees performance, on December 29, 2025.
INSPECTION RESULTS
Failure to Prescribe Instructions Appropriate to the Circumstance for Charging Activities while in Off-normal Alignment Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000370/2025004-01 Open/Closed
[H.11] -
Challenge the Unknown 71111.15 A self-revealed Green finding and NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to provide instructions or procedures for charging activities while in an off-normal CVCS alignment.
Description:
On October 13, 2025, Unit 2 was in Mode 5 (Cold Shutdown) with the reactor coolant (NC) system at approximately 140°F and the pressurizer maintaining a bubble (small steam space to help control pressure in the NC system). The plant was placed in this configuration to repair a leak on Unit 2 boric acid blender outlet relief, 2NV-483, which required the volume control tank and letdown system to be isolated. This resulted in an off-normal alignment for charging activities as part of NC system level control, with the suction source being the refueling water storage tank (FWST).
The Unit 2 train A centrifugal charging pump (2A NV) was started and aligned to deliver approximately 195 gpm of charging flow to raise pressurizer level from 30 to 50 percent. The rapid addition of cold water from the FWST, approximately 2625 gallons at 80°F, caused the pressurizer liquid space to cool rapidly, exceeding the cooldown rate limit of 200°F/hour.
Shortly after the 2A NV pump was secured, the subsequent heatup recovery resulted in exceeding the maximum heatup limit of 100°F/hour. Selected Licensee Commitment 16.5.8, Pressurizer, states that:
The pressurizer temperature shall be limited to:
1. A maximum heatup of 100°F in any 1-hour period,
2. A maximum cooldown of 200°F in any 1-hour period, and...
For the charging activity, the licensee used operating procedure OP/2/A/6200/001B, Chemical and Volume Control System Charging, Attachment 12, NV Pump Operation In Mode 5, 6, Or No Mode, with an effective revision date of October 6, 2025. However, procedure OP/2/A/6200/001B did not contain instructions of a type appropriate to the circumstance such that the charging activities could be satisfactorily accomplished while in this off-normal CVCS alignment. As a result, operations performed charging activities at flow rates that resulted in the rapid cooldown of the pressurizer fluid and exceeded the pressurizer temperature rate of change limits.
Corrective Actions: The licensee revised the applicable operating procedures as well as provided enhanced training on standards/expectations related to procedural limits, alarm response, and pre-job briefs. Accountability plans were implemented for involved personnel, and a fleet-wide operating experience evaluation was conducted. The licensee performed engineering evaluation MCC-1201.01-00-0043 to confirm the pressurizer's integrity and operability.
Corrective Action References: NCR 02571935
Performance Assessment:
Performance Deficiency: The failure to provide instructions of a type appropriate to the circumstance for charging activities affecting quality, while in an off-normal CVCS alignment, as required by 10 CFR Appendix B, Criterion V, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the exceedance of pressurizer cooldown and heatup rates resulted in reasonable doubt of the pressurizer's ability to perform its required safety function and required licensee revision to engineering calculations to establish functionality.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using IMC 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings," Exhibit 4, Barrier Integrity Screening Questions, the inspectors determined the finding to be of very low safety significance, Green, because it did not affect the reactor pressure vessel and did not result in the unavailability of a power operated relief valve or low temperature overpressure protection relief valve.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, the licensee did not ensure that all critical parameters, such as cooldown/heatup rates, were adequately addressed during pre-job briefs and execution of the evolution while in the off-normal CVCS alignment.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Contrary to the above, since at least October 6, 2025, the licensee failed to prescribe documented instructions of a type appropriate to the circumstances for an activity affecting quality. Specifically, the licensee failed to prescribe instructions to accomplish charging activities while in an off-normal CVCS alignment, which resulted in the licensee exceeding pressurizer temperature rate of change limits.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Correct Leakage of a Boric Acid Relief Valve Results in a Manual Reactor Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000370/2025004-02 Open/Closed None (NPP)71152A A self-revealed Green finding and NCV of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified when the licensee failed to correct a CAQ associated with the Unit 2 boric acid blender outlet relief valve, 2NV-483, which resulted in a four gpm NC system leak outside of containment and a manual reactor trip.
Description:
The Unit 2 boric acid blender outlet relief valve, 2NV-483, is a safety-related relief valve that provides pressure protection for the reactor makeup control subsystem of the CVCS. On September 22, 2019, the licensee documented a CAQ under work request (WR) 20153489 identifying boric acid leakage external to the upper portion of the valve body.
After initial investigation and monitoring of 2NV-483, the licensee updated the WR on January 6, 2020, noting that the boric acid leakage was recurring and repair was required to stop the leakage. The repair of 2NV-483 required placing the reactor in cold shutdown and isolating the volume control tank (VCT).
Licensee procedure AD-WC-ALL-0330, Outage Scope Determination and Control Process, 2, Outage Scope Prioritization assigns numerical priority codes to outstanding work items. Contrary to this attachment, the licensee assigned a lower priority code to the 2NV-483 repair than was appropriate for a CAQ. Consequently, the licensee failed to include 2NV-483 in outage work, or document justification for deferral in three consecutive refueling outages as required by AD-WC-ALL-0330 Attachment 3, "Scope Challenge Checklist," step 2.h, which states, "If the work has to be performed during an outage, is CAQ related, and is to be deferred past the next refueling outage, has justification for deferral been documented?"
On October 7, 2025, operators initiated WR 20295564, for an active 20 drops per minute leak from 2NV-483. Additionally, WR 20295602 was initiated for the Unit 2 NC filter outlet to VCT 3-way diversion control valve, 2NV-137A, cycling erratically (slamming into valve seat) in the low demand range while automatically controlling VCT level. The erratic behavior of 2NV-137A was previously identified via WRs in 2015 and 2018. On October 9, 2025, the licensee initiated a Unit 2 planned downpower to approximately 20 percent rated thermal power for unrelated repairs. On October 10, 2025, at 0002, operators in the auxiliary building reported a four gpm leak from 2NV-483. In response, operators entered abnormal operating procedures for rapid downpower and NC leakage within the capacity of both centrifugal charging pumps. Operators down powered to 5 percent rated thermal power and manually tripped the reactor. Operators successfully stabilized Unit 2 and contained the NC system leakage to the auxiliary building.
The licensee's root cause evaluation concluded the less than adequate risk recognition of the unresolved erratic behavior of 2NV-137A resulted in high cycle vibrational fatigue of the 2NV-483 bellows, resulting in its failure on October 10, 2025. Additionally, the root cause evaluation identified a contributing cause associated with the licensee's failure to fully implement the requirements of AD-WC-ALL-0330 for outage scoping, including prioritization and disposition of CAQ work orders, as it relates to the 2NV-483 degraded condition identified on September 22, 2019.
This violation is associated with LER 05000370/2025-002-00, "Reactor Trip, Auxiliary Feedwater Actuation, and Refueling Water Storage Tank Inoperability Due to Relief Valve Leak," which was submitted to the NRC on December 9, 2025.
Corrective Actions: Upon discovery of the 2NV-483 valve four gpm leak, the licensee responded using abnormal and emergency operating procedures to isolate the leak and downpower the unit. The 2NV-483 valve was replaced and tested satisfactorily. The licensee performed an extent of condition review and root cause evaluation and assigned several related corrective actions. The licensee also took interim actions to minimize erratic behavior of 2NV-137A and valve repair was planned for the next refueling outage.
Corrective Action References: NCR 02571809
Performance Assessment:
Performance Deficiency: The licensees failure to correct the deficient pressure retaining components on valve 2NV-483, as required by 10 CFR 50, Appendix B, Criterion XVI, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, leakage from 2NV-483 degraded to a condition that resulted in a four gpm isolable NC system leak and required a manual reactor trip.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the equipment performance attribute of the Initiating Events cornerstone. In accordance with IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 1 Initiating Events Screening Questions the inspectors determined that the issue required a detailed risk evaluation because the finding caused a reactor trip AND the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition, since the suction source for the charging pumps was lost when the VCT was isolated. A detailed risk evaluation was performed by a regional Senior Reactor Analyst (SRA) using Systems Analysis Program for Hands-On Integrated Reliability Evaluations (SAPHIRE) Version 8.2.12 and McGuire Units 1 and 2 Standardized Plant Analysis Risk (SPAR) model Version 8.83. Because the performance deficiency directly resulted in the Unit 2 plant transient that occurred on October 10, 2025, the risk was evaluated as a transient initiating event analysis in accordance with the guidance in Volume 1 of the RASP manual (Reference Section 8.4: Case 3 - Initiating Event and Mutually Inclusive SSC Unavailability). Exposure time does not apply when performing initiating event analyses. The SRA modified the SPAR model fault tree Failure of RCP Seal Injection From CVCS to include the alternate suction supply for the charging pumps (FWST) and an operator action to align FWST to charging. The failure was modeled as IE-Transient set to true and basic events CVC-MOV-OC-141 A and B (Failure of VCT Suction Path to Charging)both set to true. Credit for station FLEX equipment and procedures was also applied. The dominant accident sequence was a TRANS initiating event followed by unavailability of component cooling water and service water, and failure of operator actions to swap NC makeup to the fueling water storage tank resulting in RCP failure and core damage. The SRA determined that the incremental core damage probability for the plant transient was less than 1E06 and incremental large early release probability was less than 1E07, representing a finding of very low safety significance (Green) for McGuire Unit 2.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance. Specifically, the licensee failed to correctly prioritize, scope, and disposition a CAQ in accordance with applicable procedures on September 22, 2019, which is beyond the nominal three-year period for "present performance."
Enforcement:
Violation: 10 CFR Part 50 Appendix B, Criterion XVI states in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, between September 22, 2019 to October 10, 2025, the licensee failed to assure a condition adverse to quality associated with valve 2NV-483 was corrected.
Specifically, on September 22, 2019, the licensee identified external leakage from 2NV-483, which continued to degrade, ultimately leading to the failure of its pressure retaining parts on October 10, 2025, resulting in a four gpm NC leak and manual reactor trip.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Minor Violation 71153 Minor Violation: This minor violation is being documented as a result of inspection associated with LER 05000370/2025-001-00 and its revision 05000370/2025-001-01, "Condition Prohibited by Technical Specifications Due to Emergency Diesel Ventilation Damper Malfunction, which were submitted to the NRC on November 3, 2025, and December 4, 2025, respectively.
Technical Specification (TS) 3.8.1, AC Sources - Operating, Limiting Condition of Operation (LCO) 3.8.1.b requires, in part, that two diesel generators capable of supplying the onsite essential auxiliary power systems shall be operable while in Modes 1, 2, 3, and 4. With one 3.8.1.b diesel generator inoperable (CONDITION B), Required Action B.2, to perform surveillance requirement (SR) 3.8.1.1 for the required offsite circuit(s), has a 1-hour completion time. If the Required Actions and associated completion time of B.2, is not met (CONDITION I), then Required Action I.1 requires the unit to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Required Action I.2 requires the unit to be in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Contrary to the above, on February 16, 2025, the Unit 2, train A diesel generator was inoperable for approximately nine hours while the licensee was investigating anomalous fan motor test results on the diesel generator room ventilation system 2A1 fan. During this period, the licensee did not perform SR 3.8.1.1 within one hour in accordance with TS 3.8.1, Required Action B.2. Additionally, the licensee did not complete Required Action I.1 of condition I to be in Mode 3 within six hours after failure to perform the requirements of condition B.2.
The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensees ability to foresee and correct and therefore was not reasonably preventable. No performance deficiency was identified.
Screening: The inspectors determined the violation was minor. Surveillance Requirement 3.8.1.1 requires the licensee to verify correct breaker alignment and indicated power availability for each offsite circuit. Failure to verify power availability for each offsite circuit resulted in an issue that was administrative in nature and carried little to no safety consequence. Historical review of plant data shows that, had the licensee known the diesel generator was inoperable, SR 3.8.1.1 would have been met, and TS 3.8.1 would have allowed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or a time in accordance with risk-informed completion time for the licensee to restore the diesel generator to operable status.
Enforcement:
The licensee has taken actions to restore compliance and documented the issue in NCR 02544243. This condition prohibited by technical specifications constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On January 14, 2026, the inspectors presented the integrated inspection results to Edward Pigott and other members of the licensee staff.
THIRD PARTY REVIEWS Inspectors reviewed the Nuclear Safety Review Board report that was issued on July 8, 2025.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
OP/2/A/6350/005
AC Electrical Operation Other Than Normal Lineup
Calculations
MCC -1435.00-
00-0041
NFPA 805 Transition Risk-Informed Performance-Based Fire
Risk Evaluations
Corrective Action
Documents
Resulting from
Inspection
Nuclear Condition
Report(s)
2575267
CSD-MNS-PFP-
AB-0716-001
Auxiliary Building Elevation 716 Pre-Fire Plan
CSD-MNS-PFP-
AB-0733-001
Auxiliary Building Elevation 733 Pre-Fire Plan
CSD-MNS-PFP-
AB-0750-001
Auxiliary Building Elevation 750 Pre-Fire Plan
CSD-MNS-PFP-
DG1-0736-001
U1 Diesel Generator Building Elevation 736 Pre-Fire Plan
CSD-MNS-PFP-
DG2-0736-001
U2 Diesel Generator Building Elevation 736 Pre-Fire Plan
MC -1384-07.14-
Fire Plan Auxiliary Building Elevation 733'-00"
Fire Plans
MC -1384-07.15-
Fire Plan Aux Bldg. Elev 750+0
Corrective Action
Documents
Nuclear Condition
Report(s)
2546803, 25445617
Procedures
AD-EG-ALL-1210
Work Orders
20751399, 20732431
Miscellaneous
AD-OP-ALL-0210
Operational Risk Management: Conditional Vulnerability
Preparation Plan - WO 20751399, Replace Y0 Bushing on 2
EPA TF ATB
10/08/2025
AD-OP-ALL-0201
Protected Equipment
Procedures
OP/2/A/6100/003
Controlling Procedure for Unit Operation
209
Calculations
MCM 1201.01-
1349.001
Pressurizer Insurge/Outsurge WCAP Manuals 13588, 14718,
& 14950 (FIO)
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Nuclear Condition
Report(s)
2571813, 02570486, 02567943, 02568717
Corrective Action
Documents
Resulting from
Inspection
Nuclear Condition
Report(s)
2574138, 02577052
MCFD-1609-
04.00
Flow Diagram of Diesel Generator Starting Air System (VG)
MCFD-2554-
2.00
Flow Diagram of Chemical and Volume Control System (NV)
MCID-1499-
VG.03
Instrument Detail: D.G. Pneumatic/Hydraulic Control
Schematic
Drawings
MCTC-1609-
VG.V009-01
McGuire Nuclear Station, Units 1&2 Design Criteria,
Operability Requirements and Compensatory Measures,
Testing and Acceptance Criteria: 1/2VG-5160, 1/2VG-5161,
1/2VG-5170, 1/2VG-5171
Engineering
Evaluations
MCC-1201.01-00-
0043
Pressurizer Lower Head and Shell Qualification for
Insurge/Outsurge Events
Miscellaneous
Procedure
Revision
Request(s)
2560211
Procedures
MCS-1609.VG-
00-0001
Design Basis Specification for the VG System
Corrective Action
Documents
Nuclear Condition
Report(s)
2571809
MCFD-2554-
01.02
Flow Diagram of Chemical & Volume Control System (NV)
MCFD-2554-
2.00
Flow Diagram of Chemical & Volume Control System (NV)
Drawings
MCFD-2554-
03.00
Flow Diagram of Chemical & Volume Control System (NV)
AP/2/A/5500/04
Rapid Downpower
AP/2/A/5500/10
NC System Leakage Within the Capacity of Both NV Pumps
Procedures
OP/2/A/6100/022
Unit 2 Data Book
443
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Nuclear Condition
Report(s)
2571981
CSD-EG-MNS-
20.001
MNS IST Program Plan - 5th Interval
IP/0/A/2001/004 A
HK Air Circuit Breaker Inspection and Maintenance
IP/0/A/2001/004
H
Removal and Installation of Station Circuit Breakers
MP/0/7450/039
Control Room Chiller Refrigerant Transfer, Charging and
Leak Test
MP/0/7600/234
Removal and Installation of Flanged/Wafer Valves To Piping
MP/0/A/7450/019
Control Room Chiller Oil Pump Corrective Maintenance and
Oil Heater Replacement
MP/0/A/7600/235
Dresser Relief Valve, Flanged/Bolted Bonnet Type
Corrective Maintenance
PT/2/A/4600/031
Periodic Test Performance Verification for Mode Changes
SM/0/A/8030/001
Relief Valve Set Pressure Testing and Adjustment
Procedures
TE-MN-ALL-0202
Transformer and Apparatus Testing
Work Orders
20725603, 20722047, 20751399, 20375209, 20756340,
20730851
Corrective Action
Documents
Nuclear Condition
Report(s)
2573895
Miscellaneous
MNS Drill 25-05
10/23/2025
Corrective Action
Documents
Nuclear Condition
Report(s)
2571809
Work Orders
Work Request(s)
20003790, 20123155, 20153489, 20295564, 20295602
Corrective Action
Documents
Nuclear Condition
Report(s)
2571809, 02544243
Corrective Action
Documents
Resulting from
Inspection
Nuclear Condition
Report(s)
2571810, 02581560
AP/2/A/5500/04
Rapid Downpower
AP/2/A/5500/10
NC System Leakage Within the Capacity of Both NV Pumps
Procedures
PT/0/A/4700/045
Reactor Trip Investigation - Unit 1 Main Generator Lockout
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
PT/0/A/4700/045
Reactor Trip Investigation - Unit 2 boric Acid blender relief
valve leak
RP/0/A/5700/008
Release of Toxic of Flammable Gases
10