IR 05000313/2025004
| ML26033A438 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/03/2026 |
| From: | John Dixon NRC/RGN-IV/DORS/PBD |
| To: | Pehrson D Entergy Operations |
| References | |
| IR 2025004 | |
| Download: ML26033A438 (0) | |
Text
February 03, 2026
SUBJECT:
ARKANSAS NUCLEAR ONE - INTEGRATED INSPECTION REPORT 05000313/2025004 AND 05000368/2025004
Dear Doug Pehrson:
On December 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Arkansas Nuclear One. On January 15, 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. One of these findings involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the violation 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Arkansas Nuclear One.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 IV; and the NRC Resident Inspector at Arkansas Nuclear One. 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, John L. Dixon Jr., Chief Reactor Projects Branch D Division of Operating Reactor Safety Docket Nos. 05000313 and 05000368 License Nos. DPR-51 and NPF-6
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000313 and 05000368
License Numbers:
Report Numbers:
05000313/2025004 and 05000368/2025004
Enterprise Identifier:
I-2025-004-0002
Licensee:
Entergy Operations, Inc.
Facility:
Arkansas Nuclear One
Location:
Russellville, AR
Inspection Dates:
October 1, 2025, to December 31, 2025
Inspectors:
J. Drake, Senior Reactor Inspector
N. Greene, Senior Health Physicist
M. Mondou, Resident Inspector
A. Sanchez, Senior Project Engineer
B. Tindell, Senior Resident Inspector
E. Tinger, Acting Senior Resident Inspector
Approved By:
John L. Dixon Jr., Chief
Reactor Projects Branch D
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Arkansas Nuclear One, 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 Include Sufficient Technical Information in an Integrated Control System Work Instruction Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green FIN 05000313/2025004-01 Open/Closed None (NPP)71152A A self-revealed Green finding was identified when the licensee failed to include sufficient technical information in a work order used to calibrate function generators in the Unit 1 integrated control system. Specifically, the calibration work order did not include instructions to periodically calibrate an unused breakpoint for the loop B delta P function generator module. Therefore, the unused breakpoint was not adjusted for acceptable output response leading to an erroneous output signal causing main feedwater pump B to fail to adequately maintain steam generator water level following a reactor trip.
Failure to Maintain Transformer Fire Deluge Pull Stations Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000313/2025004-02 Open/Closed
[P.1] -
Identification 71153 A self-revealed Green finding and associated non-cited violation of Unit 1 License Condition 2.C.(8) was identified when the licensee failed to maintain in effect all provisions of the approved fire protection program. Specifically, the licensees failure to maintain manual actuation stations in the procedurally required configuration resulted in multiple spurious fire suppression deluge actuations and a Unit 1 trip. This event was reported as Licensee Event Report 05000313/2025-001-00 (ML25324A399).
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000313/2025-001-00 Automatic Reactor Trip Due to Fault on the 'B' Main Phase Transformer 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at approximately 70 percent rated thermal power during a coastdown for refueling outage 1R32. On October 10, 2025, the unit was shut down for planned refueling outage 1R32. On November 12, 2025, the reactor was made critical following completion of the refueling outage and returned to full power on November 17, 2025, where it remained for the remainder of the inspection period.
Unit 2 operated at or near full rated thermal power for the entire 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) (3 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1, spent fuel pool cooling during core offload operations during refueling outage 1R32, on October 20, 2025
- (2) Unit 1, reactor coolant system pressure, level, and temperature instruments during lowered inventory conditions during refueling outage 1R32, on October 31, 2025
- (3) Unit 1, fire water deluge systems following emergency feedwater pump room fire water deluge actuation, on December 17, 2025
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (3 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 1, north switchgear room, fire area I-2, due to safety-related equipment in the area, on October 3, 2025
- (2) Unit 1, reactor building, fire area J, due to safety-related equipment in the area, on October 7, 2025
- (3) Unit 2, volume control tank room, motor control center 2B63 room, and various other rooms, fire area HH, due to cables important to safe shutdown in the area, on October 31, 2025 71111.08P - Inservice Inspection Activities (PWR)
Due to the government shutdown, this inspection was completed remotely via in office document review. The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing documents associated with the following activities in Unit 1 during refueling outage 1R32 from December 11 to December 31, 2025.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01) (1 Sample)
The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:
- (1) Ultrasonic Examination
- 1-ISI-VE-25-003, reactor coolant, weld 12-002, suction pipe to safe-end (RCP B)
Visual 1 Examination
- 1-BOP-VT-25-004, reactor coolant, #37 reactor vessel stud, nut and flange (Stud Hole)
- 1-PT-VT-25-017 reactor coolant system P-32D, reactor coolant pump cover/bowl annulus Visual 3 Examination
- 1-ISI-VT-25-021 high-pressure injection support MU-170
- 1-ISI-VT-25-040 main steam, MS-144, constant force spring
- 1-ISI-VT-25-020 high-pressure injection, MU-120 - 1-way restraint Welding Activities
- gas tungsten arc-weld o
reactor coolant, RCP seal flow transmitters PDT-1282, FW-52, FW-53, FW-54, and FW-55, socket welds PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection
Activities (IP Section 03.02) (1 Sample)
The inspectors verified that the licensee conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately:
- (1) The licensee performed a Bare-Metal Visual Examination of the RPVCH upper surface with a robotic crawler and remote cameras and supplemented by direct visual examination in accordance with the requirements of ASME Code Case N-729-6 and as directed by CEP-NDE-0955, Visual Examination (VE) of Bare-Metal Surfaces.
One hundred percent of the required RPVCH surface was examined including 360 degrees around each CRDM nozzle annulus where it penetrates the RPVCH.
The inspector reviewed NDE report 1-ISI-VT-25-019, RVH instrumentation nozzles and vent lines.
PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)
The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:
(1)
- CZ-12A, entered into corrective action program, cleaned deposits
- CZ-15, entered into corrective action program, cleaned deposits
- CZ-17A, entered into corrective action program, cleaned deposits
- CZ-21, entered into corrective action program, cleaned deposits
- CZ-22C, entered into corrective action program, cleaned deposits
- CZ-38A, entered into corrective action program, cleaned deposits
- CZ-39, entered into corrective action program, cleaned deposits
- CZ-41A, entered into corrective action program, cleaned deposits
- CZ-42, entered into corrective action program, cleaned deposits
- DH-1003, entered into corrective action program, cleaned deposits
- DH-1409, entered into corrective action program, cleaned deposits
- DH-1449, entered into corrective action program, cleaned deposits
- DZ-13B, entered into corrective action program, cleaned deposits
- MU-22A, entered into corrective action program, cleaned deposits
- MU-2415A, entered into corrective action program, cleaned deposits
- P-32D, entered into corrective action program, cleaned deposits
- P-35B, entered into corrective action program, cleaned deposits
- PDT-1029, entered into corrective action program, cleaned deposits
- RBD-12A, entered into corrective action program, cleaned deposits
- RBD-1002A, entered into corrective action program, cleaned deposits
- RBD-1003A, entered into corrective action program, cleaned deposits
- RBD-1004A, entered into corrective action program, cleaned deposits
- RC-1053A, entered into corrective action program, cleaned deposits
- YT-78, entered into corrective action program, cleaned deposits PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04)
No ASME Code inspection of steam generator tubes was required this outage.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during Unit 2, turbine generator valve stroke testing, on October 2, 2025.
- (2) The inspectors observed and evaluated licensed operator performance in the control room during the shutdown of Unit 1 for refueling outage 1R32, on October 11, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)
- (1) The inspectors observed and evaluated Unit 2 continued training simulator activities on December 11, 2025.
- (2) The inspectors observed and evaluated Unit 1 continued training simulator activities on December 16, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Unit 1, reactor building isolation valves due to failures during stroking, on December 9, 2025
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:
- (1) Unit 1, main steam isolation valve CV-2692 stem and disk replacement, on November 20, 2025
Aging Management (IP Section 03.03) (1 Sample)
The inspectors evaluated the effectiveness of the aging management program for the following SSCs that did not meet their inspection or test acceptance criteria:
- (1) Units 1 and 2, cable jackets cracked due to elevated temperatures and age, on November 21, 2025
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Unit 1, planned elevated and high risk due to refueling outage 1R32, on October 1, 2025
- (2) Unit 1, planned elevated risk due to lowered reactor coolant system inventory, on November 13, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 1, polar crane functionality assessment due to missing hardware and controller malfunctions, on October 15, 2025
- (2) Unit 1, reactor fuel operability determination due to fuel reconstitution activities, on October 17, 2025
- (3) Unit 1, primary coolant sources outside containment functionality assessment due to leakage from the waste gas surge tank, on November 17, 2025
- (4) Unit 1, electromatic relief valve operability determination due to failure to close at low reactor coolant system pressure, on November 20, 2025
- (5) Unit 1, high-pressure injection train B operability determination due to discharge isolation valve CV-1228 failure to fully close during testing, on November 20, 2025
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Unit 1, permanent modification of emergency diesel generator B governor solenoid, on December 4, 2025
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated the Unit 1 refueling outage 1R32 activities from October 11, 2025, through November 13, 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) (3 Samples)
- (1) Unit 1, integrated engineered safeguards system test after maintenance on engineered safeguards components, on November 3, 2025
- (2) Unit 1, off-site power undervoltage and protective relay test after preventive maintenance, on November 4, 2025
- (3) Unit 1, main steam isolation valve CV-2692 test after valve rebuild, on November 20, 2025
Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) Unit 1, emergency core cooling sump inspection, on October 30, 2025
- (2) Unit 1, startup transformer No. 2 load shedding surveillance test, on November 24, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) Unit 1, service water to emergency diesel generator K-4B valve CV-3807 inservice test, on November 24, 2025
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) Unit 1, equipment hatch local leak rate test, on November 12, 2025
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
The inspectors evaluated:
- (1) Unit 2, emergency operating facility drill on failed fuel and reactor coolant system leakage outside containment with safety injection, on December 18, 2025
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)
- (1) Unit 1 (October 1, 2024, through September 30, 2025)
- (2) Unit 2 (October 1, 2024, through September 30, 2025)
MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)
- (1) Unit 1 (October 1, 2024, through September 30, 2025)
- (2) Unit 2 (October 1, 2024, through September 30, 2025)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) July 1, 2024, through September 30, 2025 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) July 1, 2024, through September 30, 2025 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) Unit 1, review of integrated control system card calibrations following failure of main feedwater to automatically control steam generator level after the September 24, 2025, reactor trip, on November 21, 2025
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02) (1 Sample)
- (1) The inspectors reviewed the licensees corrective action program for potential adverse trends that might be indicative of a more significant safety issue. The inspectors performed an in-depth review of the licensee's ability to identify fire water system leakage, evaluate each occurrence, and implement corrective actions. The inspectors determined evidence of an emerging trend in this area, verified the licensee was aware of this trend, and documented an observation. No issues of more than minor significance were identified.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000313/2025-001-00, Automatic Reactor Trip Due to Fault on the B Main Phase Transformer, (ML25324A399). The inspection conclusions associated with this LER are documented in this report as a non-cited violation under Inspection Results Section 71153. This LER is Closed.
INSPECTION RESULTS
Failure to Include Sufficient Technical Information in an Integrated Control System Work Instruction Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green FIN 05000313/2025004-01 Open/Closed None (NPP)71152A A self-revealed Green finding was identified when the licensee failed to include sufficient technical information in a work order used to calibrate function generators in the Unit 1 integrated control system. Specifically, the calibration work order did not include instructions to periodically calibrate an unused breakpoint for the loop B delta P function generator module. Therefore, the unused breakpoint was not adjusted for acceptable output response leading to an erroneous output signal causing main feedwater pump B to fail to adequately maintain steam generator water level following a reactor trip.
Description:
Following a Unit 1 reactor trip on September 24, 2025, the control room observed lowering steam generator B water level due to main feedwater pump B speed decreasing. Operators opened the feedwater pump discharge crosstie valve so main feedwater pump A could restore steam generator B water level. Investigations determined that the integrated control system (ICS), which controls main feedwater pump speed from a differential pressure error signal, was malfunctioning. Specifically, the ICS loop B delta P function generator module, which controls main feedwater pump B speed, caused an erroneous output signal post-trip due to the module input signal exceeding +10Vdc.
The function generator module normally receives one input signal and produces an output signal which approximates a curve with four straight line segments, each beginning with a breakpoint. The loop B delta P function generator receives a feedwater valve differential pressure error signal and provides an output signal that controls main feedwater pump B speed. The inspectors reviewed Model Work Order 50235594 associated with the calibration of the function generator module and determined that the fourth function generator breakpoint was listed as N/A. Vendor Manual TDB015 0710, Instruction Manual for Integrated Control and Non-Nuclear Instrumentation Systems, Revision 7, states, in part, when calibrating the function generator, the breakpoints must be determined, and all four breakpoints must be adjusted. Since the unused fourth breakpoint was not specified to be calibrated in the work order, the breakpoint was not adjusted to provide an acceptable response when the input signal was greater than +10Vdc. The licensee determined that a total of seven function generator cards for the ICS system, including the loop A delta P function generator module associated with main feedwater pump A, were adversely affected due to the calibration instructions for unused breakpoints. The loop A delta P function generator provided the correct output signal during the Unit 1 reactor trip which allowed the pump to respond properly and maintain steam generator water level.
Entergy Procedure EN-MA-106, Planning, Revision 2, Attachment 9, Work Instruction Development, states, in part, to include vendor technical information in work order instructions. Contrary to EN-MA-106, the licensee did not include the vendor technical manual requirements to periodically adjust all four function generator module breakpoints since the unused fourth breakpoint was not specified to be calibrated in the model work order.
The inadequate work instructions led to the erroneous signal which caused main feedwater pump B to reduce speed post-trip.
Corrective Actions: The licensee updated the work instruction for calibration of the seven ICS function generator modules, and the seven cards were recalibrated.
Corrective Action References: Condition Reports CR-ANO-1-2025-01383, CR-ANO-1-2025-01755
Performance Assessment:
Performance Deficiency: The licensees failure to include sufficient technical information in the ICS function generator calibration work order was a performance deficiency. The performance deficiency was within the licensees ability to foresee and correct because Procedure EN-MA-106, Planning, discusses the requirement to include vendor technical information in work instructions. Specifically, the licensee failed to plan for periodic calibrations of unused breakpoints for function generator modules, which led to main feedwater pump B being unable to maintain steam generator water level following a reactor trip.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors determined the inadequate ICS function generator calibration instructions led to main feedwater pump B reducing speed and lowering steam generator water level when feedwater was required to cool the reactor coolant system following a 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. The finding was determined to be of very low safety significance (Green) because it did not:
- (1) represent a deficiency affecting the design or qualification of a mitigating SSC;
- (2) represent a loss of the PRA function of a single train TS system;
- (3) represent an actual loss of the PRA function of one train of a multi-train TS system for more than its TS allowed outage time or;
- (4) two separate safety systems for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;
- (5) represent a loss of a PRA system and/or function as defined in the PRIB or the licensees PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and
- (6) represent a loss of a PRA function of one or more non-TS trains of equipment designated as risk-significant in accordance with the licensees maintenance rule program for greater than 3 days. Specifically, ICS is not a safety-related TS system, and the PRA functionality for the main feedwater system was maintained following the reactor trip by main feedwater pump A feeding one steam generator, ensuring continuous cooling was maintained to the reactor coolant system.
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 model work order had not been revised within the last 3 years, and there was not an opportunity to recognize the issue between the last work order update and the present.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Observation: Review of Fire Water System Leakage 71152S The inspectors performed a review of potential adverse trends in the licensees corrective action program with emphasis on fire water system leaks. Records in the sites corrective action program, in addition to other administrative records, showed there has been a negative trend of fire water leaks, including valve leakage and through-wall leakage. During a review of site condition reports over the last year, the inspectors identified 15 condition reports related to fire water valve leakage and 12 condition reports related to fire water piping through-wall leakage across Units 1 and 2. The inspectors identified two fire water leaks that were large enough to decrease fire water system pressure to approximately 110 pounds per square inch gauge (psig) and automatically start the electric fire water pump. These leaks were larger than the capacity of the jockey pump, 50 gallons per minute, which normally maintains system pressure at 118 to 134 psig. The inspectors also performed a review of the site fire impairment database and identified 11 fire impairment forms documented over the last year due to fire water system isolations caused by system leakage. Several fire impairments documented compensatory actions required, such as un-isolating valves to supply fire water and staging fire hoses. These compensatory actions can delay the site fire brigade from responding to fires.
The licensee is aware of the issues relating to the fire water system and recently documented a trend for degraded underground fire system leaks in Condition Report CR-ANO-C-2025-01688. The current state of the site fire water system could indicate the existence of a more significant safety issue caused by an aggregation of small system leaks or large ruptures stressing the system. No more than minor issues were identified.
Failure to Maintain Transformer Fire Deluge Pull Stations Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000313/2025004-02 Open/Closed
[P.1] -
Identification 71153 A self-revealed Green finding and associated non-cited violation of Unit 1 License Condition 2.C.(8) was identified when the licensee failed to maintain in effect all provisions of the approved fire protection program. Specifically, the licensees failure to maintain manual actuation stations in the procedurally required configuration resulted in multiple spurious fire suppression deluge actuations and a Unit 1 trip. This event was reported as Licensee Event Report 05000313/2025-001-00 (ML25324A399).
Description:
On September 24, 2025, Unit 1 automatically tripped from approximately 89.7 percent reactor power on a valid reactor protection system actuation due to a fault on the Unit 1 main transformer B. The cause of the actuation was determined to be a spurious actuation of the Unit 1 main transformer B deluge. The spurious actuation was caused by fire deluge hand switch HS-5636 B, a manual actuation pull station that opens UAV-5636, inadvertently opening allowing fire water to reach the transformer. HS-5636 was found with a failed tamper seal that showed signs of environmental degradation and an incorrectly sized break rod which was not securing the manual actuation station. Unit 1 Procedure OP-1104.032, Fire Protection Systems, Revision 98, includes instructions for installation of the break rod for certain manual actuation pull stations. These procedures are used during manual actuation station resetting and certain tests of the fire protection system, including the transformer deluges, and should have identified and corrected the inadequate condition of the break rods. Following the Unit 1 reactor trip, the licensee identified multiple manual actuation pull stations in Unit 1 and Unit 2 with incorrectly sized break rods. Spurious actuations caused by incorrect break rods for manual actuation pull stations installed in Unit 2 occurred on July 9, 2022, and September 23, 2025.
Corrective Actions: The licensee upgraded manual actuation pull stations on Unit 1 during a planned refueling outage; identical actions are planned for Unit 2 manual actuation pull stations.
Corrective Action References: Condition Report CR-ANO-1-2025-01382
Performance Assessment:
Performance Deficiency: The licensees failure to maintain manual actuation pull stations in the procedurally required configuration was a performance deficiency. Specifically, the licensee failed to correctly perform Procedure OP-1104.032, Fire Protection Systems, Revision 98. Specific portions of this procedure ensure manual pull stations are properly secured via a break rod and a tamper seal. The cause of the issue of concern was reasonably within the licensees ability to foresee and correct because previous spurious deluge actuations on Unit 2 main transformer A on July 9, 2022, and September 23, 2025, with manual actuation pull stations of identical design indicated the correct break rod was not installed and the tamper seal degrades over time.
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, the licensees failure to maintain manual actuation pull stations lead to a fault on the Unit 1 main transformer B and an automatic 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. The finding was determined to be of very low safety significance (Green) because it did not cause both a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Cross-Cutting Aspect: P.1 - Identification: The organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to identify the manual pull station break rods when addressing spurious deluge actuations or resetting following testing.
Enforcement:
Violation: Arkansas Nuclear One, Unit 1 License Condition 2.C.(8) states, in part, the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated January 29, 2014. The license amendment request dated January 29, 2014, Attachment A, Section 3.2.3, Procedures (1), states, in part, procedures are established for inspection, testing, and maintenance of fire protection systems.
Procedure OP-1104.032, Fire Protection Systems, Revision 98, Section 14, Resetting 4-inch and 6-inch Multimatic Deluge Valve, requires manual pull stations for UAV-5619, UAV-5635, UAV-5636, and UAV-5637 have a tamper seal and a break rod installed, with the break rod almost flush with the top of the pull station.
Contrary to the above, prior to September 24, 2025, the licensee failed to maintain in effect all provisions of the approved fire protection program, as specified in the license amendment request dated January 29, 2014, for Unit 1. Specifically, the licensee failed to maintain manual actuation pull stations with a break rod almost flush with the top of the pull station as specified in OP-1104.032, which resulted in spurious deluge actuation on the Unit 1 main transformer B on September 24, 2025, and subsequent Unit 1 automatic trip from a valid reactor protection system actuation.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On December 31, 2025, the inspectors presented the Unit 1 inservice inspection results to Gary Sullins, General Manager of Plant Operations, and other members of the licensee staff.
- On January 15, 2026, the inspectors presented the integrated inspection results to Doug Pehrson, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Drawings
M-219, Sheet 4
Piping & Instrument Diagram Deluge Valve Trim Details
Drawings
M-219, Sheet 5
Piping & Instrument Diagram Deluge Valve Trim Details
Drawings
M-232, Sheet 1
Piping & Instrument Diagram Decay Heat Removal System
111
Drawings
M-235, Sheet 1
Piping & Instrument Diagram Spent Fuel Cooling System
Procedures
OP-1015.002
Decay Heat Removal and LTOP System Control
Procedures
OP-1015.048
Shutdown Operations Protection Plan
Procedures
OP-1104.006
Spent Fuel Cooling System
Procedures
OP-1104.032
Fire Protection Systems
Corrective Action
Documents
CR-ANO-
2-2025-01499
Miscellaneous
Fire Hazards Analysis
Procedures
ANO Prefire Plan (Unit 1)
Procedures
ANO Prefire Plan (Unit 2)
Corrective Action
Documents
CR-ANO-
1-2025-01304, 1-2025-01550, 1-2025-01551, 1-2025-01557,
1-2025-01563, 1-2025-01606, 1-2025-01614, 1-2025-01620,
1-2025-01718, 1-2025-01783, 1-2025-01785, 1-2025-01837,
1-2025-01842, 1-2025-01870, 1-2025-01964, 1-2025-02131
Engineering
Changes
Pipe Support MS-144 Enhancements
NDE Reports
1-BOP-VT-25-004
- 37 Reactor Vessel Stud, Nut and Flange (Stud Hole)
10/21/2025
NDE Reports
1-ISI-VE-21-003
Reactor Vessel Head Examination
04/27/2021
NDE Reports
1-ISI-VE-25-003
Suction Pipe to Safe-End (RCP B)
10/16/2025
NDE Reports
1-ISI-VT-22-002
Cold Leg Drain Nozzle to Safe-End Circ Weld (RCP B)
10/14/2022
NDE Reports
1-ISI-VT-25-019
Reactor Vessel Head Instrumentation Nozzles and Vent Lines
10/18/2025
NDE Reports
1-ISI-VT-25-020
MU-120 - 1-Way Restraint
10/21/2025
NDE Reports
1-ISI-VT-25-021
MU-120, Variable Spring
10/22/2025
NDE Reports
1-ISI-VT-25-040
MS-144, Constant Force Spring
11/12/2025
NDE Reports
1-PT-VT-25-017
P-32D, Reactor Coolant Pump
10/23/2025
NDE Reports
Procedures
CEP-NDE-0955
Visual Examination (VE) of Bare-Metal Surfaces
308
Miscellaneous
260202
EOP-Station Blackout Simulator
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
Conduct of Operations
Procedures
OP-1102.010
Plant Shutdown and Cooldown
Procedures
OP-1102.016
Power Reduction and Plant Shutdown
Procedures
OP-1202.006
Tube Rupture
Procedures
OP-1203.023
Small Steam Generator Tube Leaks
Procedures
OP-2106.009
Turbine Generator Operations
Corrective Action
Documents
CR-ANO-
1-2008-01603, 1-2018-01458, 1-2021-01168, 1-2022-02340,
1-2025-01821, 1-2025-01840, 1-2025-02044
Miscellaneous
PO-10462164
QC Inspection for Body to Bonnet Gasket
2/22/2015
Miscellaneous
PO-62213913
Purchase Order for Disk/Stm/Stm-Dik/Piston Assy
Procedures
Safety & Quality Classification of Replacement Parts
Procedures
Procurement Engineering Process
Procedures
Cable Reliability Program
Procedures
Non-EQ Insulated Cables and Connections Inspection
Procedures
VLF Tan Delta and Withstand Testing of Electrical Power
Cables
Procedures
Motor Operated Valve Diagnostics
Work Orders
Calculations
CALC-09-E-0008-
ANO-1 NFPA 805 Non-Power Operations Assessment
Corrective Action
Documents
CR-ANO-
1-2025-01573
Miscellaneous
1R32 Outage Risk Assessment Team Report
Procedures
Protected Equipment Postings
Procedures
OP-1015.002
Decay Heat Removal and LTOP System Control
Calculations
CALC-27A
Polar Crane Girder
Corrective Action
Documents
CR-ANO-
1-2021-01589, 1-2021-01777, 1-2025-01127, 1-2025-01146,
1-2025-01571, 1-2025-01653, 1-2025-01656, 1-2025-01692,
1-2025-01725, 1-2025-01732, 1-2025-01746, 1-2025-01832,
1-2025-02085
Engineering
Changes
Provide PSV-1000 Closure Stroke Time Acceptance Criteria
Miscellaneous
PMCR-ANO-
Revise the PM Strategy of PSV-1000 IAW RV Program
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2016-00255207
Procedures
OP-1202.012
03/07/23
Procedures
OP-1305.013
Leak Testing the Gaseous Radwaste System
Engineering
Changes
K-4A&B Governor Run Solenoid Replacements
Engineering
Changes
Child EC to 0054248561 - Replace K-4B Run Solenoid
Procedures
Engineering Change Process
Work Orders
Calculations
CALC-06-E-0003-
Allowable Void Size Indication for LPI/DH Header
Corrective Action
Documents
CR-ANO-
1-2025-01368, 1-2025-02339
Miscellaneous
OpESS 2007/03
Crane and Heavy Lift Inspection, Supplemental Guidance to
09/01/2018
Procedures
Cycle 33 Core Operating Limits Report
Procedures
EN-OP-115.02
Control Room Conduct and Access Control
Procedures
EN-OP-115.14
Reactivity Management
Procedures
OP-1015.002
Decay Heat Removal and LTOP System Control
Procedures
OP-1015.048
Shutdown Operation Protection Plan
Procedures
OP-1102.002
Plant Startup
24
Procedures
OP-1102.008
Approach to Criticality
Procedures
OP-1102.016
Power Reduction and Plant Shutdown
Procedures
OP-1102.10
Plant Shutdown and Cooldown
Procedures
OP-1103.011
Draining and N2 Blanketing the RCS
Procedures
OP-1104.004
Decay Heat Removal Operating Procedure
141
Procedures
OP-1504.007
Unit 1 Reactor Vessel Closure Head Removal and Storage
Procedures
Outage Risk Assessment Team Report
Corrective Action
Documents
CR-ANO-
1-2025-01840, 1-2025-01845, 1-2025-01848
Engineering
Changes
Baseline Reference Values for ANO1 IST Components
Procedures
OP-1015.036
Containment Building Closeout
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
OP-1305.005
SU#2 Transformer Load Shedding Test
Procedures
OP-1305.006
Integrated ES System Test
Procedures
OP-1305.007
RB Isolation and Miscellaneous Valve Stroke Test
Procedures
OP-1305.037
Unit 1 Reactor Building Access & Ventilation Leak Rate
Testing
Procedures
OP-1305.059
Off-site Power Undervoltage and Protective Relay Test
Procedures
OP-1402.038
Unit 1 Equipment Hatch Opening, Closing, and Maintenance
Procedures
OP-1402.069
Main Steam Isolation Valve (MSIV) Disassembly, Inspect.,
Repair & Reassembly
Work Orders
WO 54141816, 54165044, 54167396, 54169938, 54174197,
210679
Miscellaneous
Red Team 4th Qtr PI Drill
2/18/2025
71151
Calculations
CALC-ANO1-SA-
16-00001
ANO-1 Mitigating System Performance Index Basis
71151
Calculations
CALC-ANO2-SA-
06-00001
ANO-2 Mitigating System Performance Index Basis
71151
Procedures
Regulatory Performance Indicator Process
71151
Self-Assessments
LO-ANO-2024-
00052
Radiation Inspection - Performance Indicator Verification -
07/09/2025
Corrective Action
Documents
CR-ANO-
1-2025-01755, 1-2025-02153
Engineering
Changes
Expanding Calibration of Function Generators with Unused
Break Points
Work Orders
Corrective Action
Documents
CR-ANO
C-2025-01688
Procedures
OP-1000.120
ANO Fire Impairment Program
Corrective Action
Documents
CR-ANO-
1-2025-01382
Procedures
OP-1104.032
Fire Protection Systems
98