IR 05000313/2024004
| ML25029A062 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/31/2025 |
| From: | John Dixon NRC/RGN-IV/DORS/PBD |
| To: | Pehrson D Entergy Operations |
| Sanchez A | |
| References | |
| IR 2024001 | |
| Download: ML25029A062 (33) | |
Text
January 31, 2025
SUBJECT:
ARKANSAS NUCLEAR ONE - INTEGRATED INSPECTION REPORT 05000313/2024004 AND 05000368/2024004 AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION REPORT 07200013/2024001
Dear Doug Pehrson:
On December 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Arkansas Nuclear One. On January 16, 2025, 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. Also, one Severity Level IV violation without an associated finding is documented in this report. 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Arkansas Nuclear One.
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 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; 05000368; 07200013 License Nos. DPR-51; NPF-6
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000313; 05000368; 07200013
License Numbers:
Report Numbers:
05000313/2024004; 05000368/2024004; 07200013/2024001
Enterprise Identifier:
I-2024-004-0000; I-2024-001-0117
Licensee:
Entergy Operations, Inc.
Facility:
Arkansas Nuclear One and ISFSI
Location:
Russellville, AR
Inspection Dates:
October 1, 2024, to December 31, 2024
Inspectors:
C. Alldredge, Health Physicist
C. Borman, Health Physicist
L. Brookhart, Senior Spent Fuel Storage Inspector
T. DeBey, Resident Inspector
J. Drake, Senior Reactor Inspector
J. Freeman, Spent Fuel Storage Inspector
A. Sanchez, Senior Project Engineer
B. Tindell, Senior Resident Inspector
E. Tinger, 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 Ensure HI-STORM FW Version E1 Design Change did not Require a Change to the Certificates Technical Specifications Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000313,05000368/2024004-01 Open/Closed Not Applicable 60855 The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 72.48(c)(1)(ii)(B)for the licensees failure to request that the certificate holder (Holtec) obtain a Certificate of Compliance (CoC) amendment pursuant to 10 CFR 72.244, prior to implementing a proposed design change that required a change to a technical specification.
Failure to Follow Procedure for Control of Combustible Liquids Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000368/2024004-02 Open/Closed
[H.4] -
Teamwork 71111.08P The inspectors identified a Green finding and associated non-cited violation of License Condition 2.C.(3)(b), Fire Protection, for Arkansas Nuclear One Unit 2, associated with the failure to adequately implement the fire protection program. Specifically, the licensee failed to follow the requirements for control of combustibles for approximately 12 Just Rite safety cans containing flammable liquids in the hot tool room inside the radiologically controlled area in the plant. On October 3, 2024, the licensee removed the containers and stored them in an approved location. The licensee entered these issues into their corrective action program as Condition Report CR-ANO-2-2024-02424.
Failure to Properly Evaluate Operability of the Unit 2 Service Water Intake Structure After Damage from Acid Exposure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000368/2024004-03 Open/Closed
[H.11] -
Challenge the Unknown 71111.15 The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete an adequate operability determination in accordance with Procedure EN-OP-104, Operability Determination Process, revision 19. Specifically, the licensee failed to appropriately evaluate the operability of the intake bay structure and related safety components for Unit 2 service water pump 2P-4A after damage from long-term acid exposure, which is a performance deficiency.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000313,05000368/2023004-04 Holtec HI-STORM FW Overpack Version E1 Vent Design Change 60855 Closed LER 05000368/2024-001-00 LER 2024-001-00 for Arkansas Nuclear One,
Unit 2, Surface Flaw in Primary Coolant System 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at full power. On December 4, 2024, the unit was shut down to Mode 5 to repair a leaking pressurizer code safety valve. Upon completion of repairs, operators restarted Unit 1 on December 9, 2024, and the unit was returned to full power on December 10, 2024. Unit 1 remained at or near full power for the remainder of the inspection period.
Unit 2 began the inspection period in refueling outage 2R30. Following the refueling outage, Unit 2 was restarted on October 28, 2024, and returned to full power on October 31, 2024.
Unit 2 remained at or near full power 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
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures for the following systems:
- Units 1 and 2 service water intake structure, Unit 1 emergency feedwater initiation and control pressure sensing lines, and the Units 1 and 2 alternate ac diesel generator, on November 5, 2024
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 2, reactor coolant system level and temperature instrumentation during reduced reactor coolant system inventory, on October 1, 2024
- (2) Unit 2, shutdown cooling system during refueling outage 2R30, on October 17, 2024
- (3) Unit 1, decay heat removal system during the planned shutdown, on December 6, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (2 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, containment, FZ-2001, FZ-2012, FZ-2013, FZ-2014, FZ-2015, on October 3, 2024
- (2) Unit 2, turbine building, FZ-2200-MM, on October 28, 2024
===71111.08P - Inservice Inspection Activities (PWR) 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 the following activities in Unit 2 during refueling outage 2R30 from September 30, 2024, to November 26, 2024.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===
The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:
- (1) Ultrasonic Examination
- reactor coolant, "B" cold leg drain line, component 36-003, elbow to pipe weld
- reactor coolant, "C" cold leg drain line, component 38-003, elbow to pipe weld
- reactor coolant, "D" cold leg drain line, component 39-003, elbow to pipe weld
- reactor coolant, "D reactor coolant pump, component 39-001, nozzle to safe end weld (phased array)
Visual Examination
- high-pressure safety injection, variable spring 2HCB-15-H7
- reactor coolant system,39-001, nozzle to safe end circumferential weld Welding Activities
- gas tungsten arc welding o
main steam, 2MS-2082A/B, FW-4, FW-5, FW-6, FW-7, FW-8, FW-9 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)
- ultrasonic examination of penetrations 1 - 73
- eddy current examination of penetration 46, 48, 49, 53, 67, 71
- electronic visual examination of penetrations 35, 43, 61, 67, 71, and 78
- relevant indication identified on penetration 71 and half-nozzle repair completion 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)
- Condition Report ANO-2-2020-03059
- Condition Report ANO-2-2022-01016
- Condition Report ANO-2-2023-00296
- Condition Report ANO-2-2023-00627
- Condition Report ANO-2-2023-00628
- Condition Report ANO-2-2023-00630
- Condition Report ANO-2-2023-00632
- Condition Report ANO-2-2023-00700
- Condition Report ANO-2-2023-00744
- Condition Report ANO-2-2023-01670
- Condition Report ANO-2-2023-01697
- Condition Report ANO-2-2023-01957
- Condition Report ANO-2-2023-01961
- Condition Report ANO-2-2023-02610
- Condition Report ANO-2-2023-02048
- Condition Report ANO-2-2023-02050
- Condition Report ANO-2-2023-02051
- Condition Report ANO-2-2023-02359
- Condition Report ANO-2-2023-02464
- Condition Report ANO-2-2023-02465
- Condition Report ANO-2-2023-02468
- Condition Report ANO-2-2023-02472
- Condition Report ANO-2-2023-02608
- Condition Report ANO-2-2023-02609
- Condition Report ANO-2-2024-01080
- Condition Report ANO-2-2024-01145
- Condition Report ANO-2-2024-01166 PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04)
No steam generator tube inspection was required this refueling 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 the Unit 2 approach to critical and low power operations after refueling outage 2R30, on October 28, 2024.
- (2) The inspectors observed and evaluated licensed operator performance in the control room during the Unit 1 approach to critical and low power operations after the shutdown, on December 9, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)
- (1) The inspectors observed and evaluated Unit 1 simulator evaluation and continued training, on November 13, 2024.
- (2) The inspectors observed and evaluated Unit 2 simulator evaluation and continued training, on November 21, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (4 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) Unit 2, steam generator level instrumentation, on November 20, 2024
- (2) Unit 2, plant protection system, on November 20, 2024
- (3) Unit 1, decay heat removal system, on November 25, 2024
- (4) Unit 2, radiation area monitors, on December 12, 2024
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 2, emergency diesel generator fuel injector and potential transformer fuse door mechanism replacement, on November 6, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (4 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 2, elevated risk for work on startup 2 transformer during refueling outage 2R30, on October 2, 2024
- (2) Units 1 and 2, elevated risk during startup transformers 1 and 3 outages, on October 8, 2024
- (3) Unit 1, elevated risk for service water pump P-4C and electric-driven fire pump P-6A intake bay outage, on November 19, 2024
- (4) Unit 1, elevated risk due to venting reactor coolant system during planned shutdown, on December 6, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 2, operability determination of reactor coolant system loops due to flow restriction, on November 5, 2024
- (2) Unit 2, operability determination of reactor coolant system flow measurement due to flow reduction, on November 6, 2024
- (3) Unit 2, operability determination of emergency diesel generator 2K-4B after speed signal oscillations, on November 17, 2024
- (4) Unit 2, operability determination of service water intake bay A after acid damage, on November 21, 2024
- (5) Unit 2, operability determination of service water pump 2P-4B after pump maintenance, on December 17, 2024
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)
- (1) The inspectors evaluated Unit 2 refueling outage 2R30 activities from October 1, 2024, through October 28, 2024. The inspectors completed inspection procedure sections 03.01.c through 03.01.e.
- (2) The inspectors evaluated Unit 1 planned outage activities from December 4, 2024, through December 9, 2024.
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) (5 Samples)
- (1) Unit 2, emergency diesel generator exhaust fans 2VEF-24A and 2VEF-24B post-maintenance test after maintenance, on November 19, 2024
- (2) Unit 2, atmospheric dump isolation valve 2CV-1001 post-maintenance test after outage maintenance, on November 21, 2024
- (3) Unit 2, main steam isolation valve 2CV-1010 post-maintenance test after replacement, on October 29, 2024
- (4) Units 1 and 2, common feedwater post-maintenance test after maintenance on FW-2847, on December 11, 2024
- (5) Unit 2, downstream atmospheric dump valve 2CV-0305 post-maintenance test after maintenance, on December 12, 2024
Surveillance Testing (IP Section 03.01) (3 Samples)
- (1) Unit 2, control element assembly drop time testing, on October 29, 2024
- (2) Unit 2, emergency systems response time testing, on October 24, 2024
- (3) Unit 2, emergency-powered pressurizer heater capacity testing, on November 6, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) Unit 2, high-pressure safety injection pump 2P-89C in-service testing, on November 15, 2024
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) Unit 2, local leak rate testing of the high-pressure nitrogen to safety injection tank containment isolation valve 2CV-6207-2, on October 8, 2024
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
- (1) Units 1 and 2, FLEX steam generator/reactor coolant system makeup pump P-255 flow test, on December 10,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
(1)licensee surveys of potentially contaminated material leaving the radiologically controlled area (RCA)
(2)workers exiting the RCA at Unit 2 during a refueling outage
Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
(1)radiological work permit (RWP) 20242471, 2R30 inspections on Unit 2 reactor head on October 8, 2024
- (2) RWP 20242471, 2R30 removal of genesis robot from under Unit 2 reactor head on October 9, 2024
- (3) RWP 20242407, decontamination of genesis robots on October 9, 2024
- (4) RWP 20242433, work on upper guide structure lifting rig work platform on October 10, 2024 High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (3 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs):
- (1) HRA Unit 2, underneath reactor vessel head on the reactor building 404-foot elevation
- (3) HRA Unit 2, auxiliary building 335-foot elevation 2T-5 tank room Radiation Worker Performance and Radiation Protection Technician Proficiency (IP
Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Permanent Ventilation Systems (IP Section 03.01) (1 Sample)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
- (1) Unit 2 reactor building ventilation
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) Unit 2 high-efficiency particulate air ventilation for under reactor vessel head inspection
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
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)===
- (1) Unit 1 (October 1, 2023, through September 30, 2024)
- (2) Unit 2 (October 1, 2023, through September 30, 2024)
MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)
- (1) Unit 1 (October 1, 2023, through September 30, 2024)
- (2) Unit 2 (October 1, 2023, through September 30, 2024)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) Units 1 and 2 (April 1, 2023, through June 30, 2024)
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) Units 1 and 2 (April 1, 2023, through June 30, 2024)
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Unit 2, effectiveness of corrective actions for repeat failures of the latch on the emergency diesel generator 2K-4B potential transformer fuse cabinet door, on November 7, 2024
- (2) Unit 2, effectiveness of corrective actions related to a failure of turbine-driven emergency feedwater pump steam supply check valve 2MS-39B, on December 11, 2024
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program to identify potential trends with emphasis on equipment issues tracked by the operations department and maintenance backlog lists that might be indicative of a more significant safety issue.
The inspectors determined evidence of a trend in this area, informed the licensee 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)
The inspectors evaluated the following licensees event reporting determinations to ensure it complied with reporting requirements.
- (1) LER 05000368/2024-001-00, Surface Flaw in Primary Coolant System (ADAMS Accession No. ML24338A212). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensee's ability to foresee and correct, and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified. This LER is Closed.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
===60855 - Operation Of An ISFSI The inspectors performed a review to close an unresolved item regarding the licensees independent spent fuel storage installation (ISFSI) activities.
Operation Of An ISFSI===
- (1) In accordance with Inspection Procedure 60855, section 04.03, inspectors performed an in-office follow-up review of unresolved item (URI) 05000313,05000368/2023004-04. This review identified a violation for ANO's failure to request Holtec obtain a Certificate of Compliance amendment prior to implementing the HI-STORM FW Version E1 design change that required a change in technical specifications.
INSPECTION RESULTS
Failure to Ensure HI-STORM FW Version E1 Design Change did not Require a Change to the Certificate's Technical Specifications Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000313,05000368/2024004-01 Open/Closed Not Applicable 60855 The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 72.48(c)(1)(ii)(B)for the licensees failure to request that the certificate holder (Holtec) obtain a Certificate of Compliance (CoC) amendment pursuant to 10 CFR 72.244, prior to implementing a proposed design change that required a change to a technical specification.
Description:
In NRC inspection report titled "Arkansas Nuclear One - Integrated Inspection Report 05000313/2023004, 05000368/2023004 and ISFSI Inspection Report 07200013/2023002, " dated February 8, 2024 (ADAMS Accession No. ML24039A084), the inspection team opened an unresolved item (URI)
No. 05000313,05000368/2023004-04 regarding a design change that ANO had adopted and chose to utilize during their first loading campaign associated with the HI-STORM FW system. This design change moved the inlet vents on the FW overpacks higher, above ground level, to preclude floodwater ingress into the canister and installed a drain plug at the bottom of the cask system. Changing the height of the inlet vents created a potential to trap water related to floods or rain in the system. If enough water entered the vents to block airflow, this would result in an adverse thermal effect on the canister. Floodwater entering the system was evaluated by the licensee through a thermal evaluation. As part of that thermal evaluation, the licensee lowered the design-basis heat load to ensure Final Safety Analysis Report (FSAR) limitations on short-term peak fuel cladding temperatures would remain below design-basis limits. At the time, the inspectors questioned the licensee if rainwater could enter through either the inlet or outlet vents. This scenario would be similar to the floodwater ingress event; however, it would be undetectable when performing the daily technical specification vent surveillance to ensure vents remain unblocked. The inspectors also learned of operating experience from other licensees confirming rainwater ingress had occurred at multiple sites utilizing this FW overpack design. The inspectors subsequently submitted a technical assistance request (TAR) to the Division of Fuel Management to evaluate if rainwater should have been accounted for and if this design change met 10 CFR 72.48 requirements.
The NRC determined this design change did not meet the requirements of 10 CFR 72.48.
Certificate of Compliance 72-1032 HI-STORM FW Storage System Appendix A, Technical Specification 3.1.2 surveillance requirement states: "Verify all overpacks inlets and outlets are free of blockage from solid debris or floodwater at a frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." The technical bases in Appendix 13.A of the FW FSAR, TS Limiting Condition of Operation (LCO) 3.1.2, stated, the TS is not intended to address low frequency, unexpected Design Event III and IV class events (ANSI/ANS-57.9) such as design-basis accidents and extreme environmental phenomena that could potentially block one or more of the air ducts for an extended period...the intent of this LCO is to address occurrences of air blockage that can be reasonably anticipated to occur from time to time at the ISFSI (i.e., Design Event I and II class events per ANSI/ANS-57.9). These events are of the type where corrective actions can usually be accomplished within one 8-hour operating shift to restore the heat removal system to operable status (e.g., removal of loose debris)."
The NRC determined that the location of the inlet vents would make it difficult for the licensees personnel to perform a visual observation that detects the presence of water trapped in the HI-STORM FW Version E1, unless the water was pouring out of the inlet vents.
The inlet and outlet vents are located well above the ground level where rainwater may enter the overpack and remain trapped without being visually observed. The trapped water if not drained would interfere with the air flow and cause the fuel assemblies stored in the multi-purpose canister (MPC) to exceed design-basis temperature limits unless restrictions were in place to limit the decay heat load of the canister. Additionally, as stated in the TS Bases, it is assumed under normal storage conditions, the inlet and outlet air ducts are unobstructed and allow full air flow." Since the design change limits the draining of the water within the overpack (unless the plug is removed), it could impede full air flow and it would be difficult to detect whether water is present at the bottom of the overpack. Therefore, the licensee may not know that it is necessary to perform Technical Specification Surveillance Requirement 3.1.2 requirement to remove the debris.
Corrective Actions: The licensee placed the issue into the corrective action program and removed the drain plugs for the three in service HI-STORM FW canisters on the ISFSI pad.
This action ensures Technical Specification 3.1.2 regarding canister heat removal vent surveillances is unchanged by establishing a path for water to drain from the anulus region of the canister preventing internal blockage. The licensee is currently evaluating the use of meshed foreign material exclusion (FME) covers on the drain ports to preclude future blockage of this drain path. The proposed new meshed FME covers would continue to allow water to exit the system.
Corrective Action Reference: CR-ANO-C-2023-03873
Performance Assessment:
None. The Reactor Oversight Process (ROP) does not specifically consider violations of 10 CFR Part 72 in its assessment of licensee performance. In accordance with Manual Chapter 0612, Appendix B, Issue Screening Directions, this violation follows the traditional enforcement path and is assessed in accordance NRCs Enforcement Policy.
Enforcement:
The ROP's significant determination process (SDP) does not address 10 CFR Part 72 violations; therefore, it is necessary to address this violation using traditional enforcement. This violation was dispositioned per the traditional enforcement process using section 2.3 of the NRCs Enforcement Policy. Traditional enforcement violations are not assessed for cross-cutting aspects.
Severity: Consistent with guidance in the NRC Enforcement Manual, Part 1, section 1.2.6.D, if a violation does not fit an example in the Enforcement Policy violation examples, it should be assigned a severity level:
- (1) commensurate with its safety significance; and
- (2) informed by similar violation addressed in the violation examples. The inspectors determined that the violation was similar to section 6.1.d.2 of the Enforcement Policy as a Severity Level IV violation. The violation was of very low safety significance due to ANO having placed a heat load limit on the casks loaded at their site. This limit ensured if water had been trapped in the system, the fuel's peak cladding temperatures would have remained below design-basis limits; therefore, not increasing the consequence of an accident or creating a possibility for a malfunction of structure, system, or component with a different result than previously evaluated.
Violation: Title 10 CFR 72.48(c)(1)(ii)(B) requires, in part, a licensee may make changes in the cask design without obtaining a Certificate of Compliance (CoC) amendment submitted by the certificate holder pursuant to 72.244 if a change to a technical specification is not required.
Contrary to the above, from September 2023 to November 2024, ANO failed to request that the certificate holder obtain a CoC amendment, prior to implementing a proposed change that required a change to a technical specification. Specifically, ANOs 10 CFR 72.48 #1600 revision 0, adopted a design change that raised the height of the inlet vents which allowed the potential for rainwater to become trapped within the system and cause the licensee the inability to adequately perform Technical Specification 3.1.2 requirements to verify the inlet vents are free of blockage at the required frequency.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
The disposition of this violation closes URI: 05000313,05000368/2023004-04.
Unresolved Item (Closed)
Holtec HI-STORM FW Overpack Version E1 Vent Design Change URI 05000313,05000368/2023004-04 60855
Description:
In September 2023, the licensee loaded their first HI-STORM FW Version E1 overpack. This new overpack design moved the inlet vents higher, above ground level, to preclude floodwater ingress into the canister. Changing the height of the inlet vents creates a potential trap for floodwater. If enough floodwater enters the vents to block airflow, this could result in an adverse thermal effect on the MPC. This scenario was evaluated by the licensee through a thermal evaluation. The inspectors questioned the licensee if rainwater can enter through either the inlet or outlet vents. This scenario would be similar to the floodwater ingress event; however, it would be undetectable when performing the daily technical specification vent surveillance. The inspectors learned of operating experience from other licensees that rainwater ingress has occurred at multiple sites utilizing other FW overpack designs. The inspectors have submitted a TAR to the Division of Fuel Management to evaluate if rainwater ingress is possible and if the thermal analysis provided for such an event is adequate.
The inspectors documented a Severity Level IV, non-cited violation in Report Section 60855 of this report.
Corrective Action Reference: CR-ANO-C-2023-03873 Failure to Follow Procedure for Control of Combustible Liquids Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000368/2024004-02 Open/Closed
[H.4] -
Teamwork 71111.08P The inspectors identified a Green finding and associated non-cited violation of License Condition 2.C.(3)(b), Fire Protection, for Arkansas Nuclear One Unit 2, associated with the failure to adequately implement the fire protection program. Specifically, the licensee failed to follow the requirements for control of combustibles for approximately 12 Just Rite safety cans containing flammable liquids in the hot tool room inside the radiologically controlled area in the plant. On October 3, 2024, the licensee removed the containers and stored them in an approved location. The licensee entered these issues into their corrective action program as Condition Report CR-ANO-2-2024-02424.
Description:
During plant walkdowns between September 30 and October 3, 2024, the inspectors observed approximately 12 "Just Rite safety cans containing flammable liquids stored on open shelving in the hot tool room inside the radiologically controlled area in the plant. Procedure EN-DC-330, Fire Protection Program, revision 4, section 2, Fire Prevention and Protection, item
- (a) stated that, Control of transient combustibles shall be implemented by administrative controls to govern the handling, storage, and limitations for use of ordinary combustible materials, combustible and flammable gases and liquids, and other combustible supplies in accordance with EN-DC-161. Procedure EN-DC-161, Control of Combustibles, revision 26, section 7.3.a, requires that flammable liquids be stored in approved (UL Listed / Factory Mutual Approved) flammable liquid storage cabinets when not in use.
Corrective Actions: The licensee entered the issue into their corrective action program and removed the "Just Rite" safety cans from the hot tool room.
Corrective Action References: Condition Reports CR-ANO-2-2024-01692 and CR-ANO-2-2024-02424
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to implement all provisions of the approved fire protection program was a performance deficiency. Specifically, the licensees failure to store combustible liquids in the reactor building hot tool room in approved flammable liquid storage cabinets in accordance with Procedure EN-DC-161, Control of Combustibles, revision 26, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors 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 performance deficiency resulted in an increased risk of fire in the radiologically controlled area of Unit 2.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix F, Fire Protection and Post - Fire Safe Shutdown SDP. The finding was determined to be within the Fire Prevention and Administrative Controls area using attachment 1 and attachment 2 as having a high degradation rating for failing to ensure the proper storage of combustible/flammable liquids. The inspectors assessed the significance of the finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix F, attachment 1, and determined the finding to be of very low safety significance (Green) because the fire adversely affected an area with adequate automatic detection and suppression.
Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Work groups and individuals were not communicating and coordinating activities within and across organizational boundaries to ensure nuclear safety is maintained and it resulted in the improper storage of flammable/combustible liquids in the hot tool room.
Enforcement:
Violation: License Condition 2.C.(3)(b), Fire Protection, requires that written procedures be established, implemented, and maintained covering fire protection program implementation.
Procedure EN-DC-330, Fire Protection Program, revision 4, section 2, Fire Prevention and Protection, requires that control of combustible materials shall be implemented by administrative controls to govern the handling, storage, and limitations for use of ordinary combustible materials, combustible and flammable gases and liquids, and other combustible supplies in accordance with Procedure EN-DC-161, Control of Combustibles.
Procedure EN-DC-161, revision 26, section 7.3.a requires that flammable liquids be stored in approved (UL Listed / Factory Mutual Approved) flammable liquid storage cabinets when not in use.
Contrary to the above, prior to October 3, 2024, the licensee failed to follow Procedure EN-DC-161 to store flammable liquids in approved (UL Listed / Factory Mutual Approved) flammable liquid storage cabinets when not in use. Specifically, inspectors identified approximately 12 "Just Rite safety cans containing flammable liquids stored on open shelving in the hot tool room inside the radiologically controlled area in the plant.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
Failure to Properly Evaluate Operability of the Unit 2 Service Water Intake Structure After Damage from Acid Exposure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000368/2024004-03 Open/Closed
[H.11] -
Challenge the Unknown 71111.15 The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete an adequate operability determination in accordance with Procedure EN-OP-104, Operability Determination Process, revision 19. Specifically, the licensee failed to appropriately evaluate the operability of the intake bay structure and related safety components for Unit 2 service water pump 2P-4A after damage from long-term acid exposure, which is a performance deficiency.
Description:
On May 29, 2024, the licensee performed a surveillance test of service water pump 2P-4A and the Unit 2 sluice gate to the emergency cooling pond (ECP) 2CV-1471-1 by aligning service water to the ECP. Following the surveillance test, 2CV-1471-1 was closed but later operators noticed that the level of the ECP was decreasing. Further checks showed that 2CV-1471-1 had significant leak by and the ECP was being drained to the intake structure. The licensee re-aligned the intake bay for service water pump 2P-4A to the ECP, which stopped the water loss. Subsequent testing showed the leak rate through sluice gate 2CV-1471-1 was about 750 gallons per minute if the bay was aligned to the lake. Divers were sent into the bay for service water pump 2P-4A, and they discovered that there were rocks on 2CV-1471-1 that had come from the intake structure concrete wall area above the valve and had prevented the sluice gate from making a good seal.
The licensee identified that the rocks had come from service water intake structure concrete wall degradation and negatively affected the closure of 2CV-1471-1. However, the rocks had not prevented the opening of the valve (its safety function). Condition Report CR-ANO-2-2024-00809 was written and it stated that the concrete wall degradation was the outer surface and is not jeopardizing the structural integrity of the wall. The focus of the operability determination was the potential for the loose rocks to enter the service water pump suction. However, the inspectors noted that the site did not document why rocks were falling onto the sluice gate, the depth of the concrete damage, or if other components were damaged.
The inspectors reviewed the issue, had discussions with numerous licensee personnel, and viewed the video from the divers inspection of the intake bay. The inspectors noted the following:
1. The potential for concrete wall debris to get stuck in the sluice gate wedge assembly
or gate track area, damage the sluice gate, and possibly prevent the gates open safety function was not discussed in the operability determination.
2. The concrete wall damage was described as affecting the outer surface, but
aggregate rocks were coming out of the concrete wall such that the inspectors determined that the damage was more than superficial.
Further questioning by the inspectors led the site to identify that the concrete wall damage was being caused by the pumping of a chemical additive (phosphoric acid) that was unintentionally spraying onto the concrete wall above the water level due to a configuration change in the chemical tubing that was made in early 2021. Prior to that change, the chemical was being pumped into the bay water below the water level through seismically mounted tubing. The phosphoric acid was also flowing onto two cast iron structural supports and anchor bolts for the stem shaft of the sluice gate, degrading those shaft supports and potentially affecting valve operation.
The inspectors questioned the operability determination for the acid damage since it was still occurring, the licensee had not fully inspected the damage, and not all negative factors had been considered.
Subsequently, the licensee moved the modified chemical discharge tube on August 19, 2024, so that the phosphoric acid was no longer being pumped onto the concrete wall. A work scope addition was made to the fall refueling outage to inspect the intake bay wall, stem shaft supports, and the sluice gate to assess the acid-caused degradation and to help plan the concrete wall repair.
Another condition report was written, and an additional operability determination was performed on August 22, 2024, with the licensee concluding that based on the minimal tolerances associated with the seating surfaces and the orientation of the sluice gate structure, it is unlikely that valve travel in the open direction will be challenged by debris. A reasonable expectation of operability is maintained regarding open travel for 2CV-1471-1.
The licensee also said that the stem shaft supports were made of materials that should not experience significant damage from the acid, but an evaluation was not done for the potential loosening of the wedge anchors that hold the shaft supports to the concrete wall. Damage to the concrete wall was assessed as not of significance from a structural standpoint. The material that is removed is the outer layer of the concrete which has exposed small aggregate rocks that have come loose over time. From the pictures, the loss of material as compared to the unaffected areas immediately adjacent to the impacted area is estimated to be approximately 1/4 to 1/2 inch depth.
Detailed inspections during the fall outage in October 2024 determined that the loss of material from the concrete wall was up to 2 inches deep, which was the same depth for the placement of rebar in the concrete and much deeper than the prior operability determinations had assumed. No rebar was visible in the damaged area, but continued acid exposure would have made that likely. It was also determined that one of the wedge blocks for the sluice gate had been bent, likely due to concrete debris being present on the wedge face. That bent wedge block was negatively affecting the seal of the sluice gate, and the licensee took the corrective action of replacing it during the unit outage. There was also a layer of concrete debris on the sluice gate that was cleaned off during the outage inspection. Visible degradation of a cast iron sluice gate stem shaft support was also noted. The licensee concluded that the degraded structure continues to have adequate structural capacity, but future nondestructive testing and repair of the wall needs to be done. A tracking action was initiated for those activities. A third operability determination was performed with consideration of the October 2024 inspection results and determined that the sluice gate was operable because it could perform its open safety function.
The inspectors determined that the licensee did not complete an adequate operability determination as directed by fleet Procedure EN-OP-104, Operability Determination, after concrete debris was found to be negatively affecting the operation of the sluice gate on May 29, 2024. Without questioning from the NRC inspectors, the acid damage and equipment degradation would still be continuing and there could be a challenge to the assumption of operability. Condition Reports CR-ANO-2-2024-02295 and CR-ANO-2-2024-02296 specifically address the inadequate operability determination.
Corrective Actions: The licensee entered this issue in their corrective action program, rerouted the chemical discharge tube in August 2024, and performed a detailed inspection of the concrete damage in October 2024.
Corrective Action References: Condition Reports CR-ANO-2-2024-00809, CR-ANO-C-2024-01126, CR-ANO-2-2024-01195, CR-ANO-2-2024-02295, CR-ANO-2-2024-02296; and Work Orders WO 54149261, WO 54094397, along with tracking action WT-ANO-2024-00282-00001
Performance Assessment:
Performance Deficiency: The licensees failure to appropriately evaluate and document the operability of the Unit 2 intake bay structure for service water pump 2P-4A and related safety components after damage from long-term acid exposure is a performance deficiency.
Specifically, the licensee failed to fully investigate and evaluate ongoing corrosion of the structural concrete, fasteners, and supports for service water components from continuing phosphoric acid injection intended for treating service water.
Screening: The inspectors determined the performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate the damage occurring in the service water intake bay would have resulted in continuing acid exposure that would further adversely impact the structural wall and sluice gate supports, as well as potential damage to the sluice gate from concrete debris.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, dated January 1, 2021. The finding was determined to be of very low safety significance (Green) because it:
- (1) did not represent a deficiency affecting the design or qualification of a mitigating SSC;
- (2) did not represent a loss of the PRA function of a single train TS system;
- (3) did not 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) did not 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) did not 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.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, the site performed intake bay inspections in April 2023 without identifying the damage and when the site first encountered the problem of the sluice gate not operating correctly and identified loose rocks and concrete debris on the gate, workers did not stop to evaluate these uncertain conditions so that the cause and risk could be known and managed.
Enforcement:
Violation: Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, 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.
Procedure EN-OP-104, Operability Determination Process, revision 19, a quality-related procedure, provides instructions in sections 7.1 and 7.2 on how to evaluate operability and requires determining whether the component is capable of performing its specified safety function.
Procedure EN-OP-104, section 5.0, step 3 states, in part, adequate documentation is necessary to establish a basis to allow for subsequent independent reviews. Operability determinations are documented in enough detail to allow an individual knowledgeable in the technical discipline associated with the condition to understand the basis for the determination. Step 4 of section 7.1 also states that a condition that could affect other components is corrosive substance leakage, and that the effect on other components should be evaluated if that exists.
Contrary to above, from May 29, 2024, through October 31, 2024, the licensee failed to accomplish activities affecting quality in accordance with procedures. Specifically, the licensee failed to evaluate operability in accordance with Procedure EN-OP-104. The licensee did not question the concrete damage to identify the acid spray on the intake bay for service water pump 2P-4A, nor did they evaluate the effect the acid might have on other components. Specifically, the effects of acid exposure creating loose rocks and other concrete debris, along with valve shaft support degradation and gate wedge damage, were not appropriately evaluated and documented in the licensees operability determination.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
Observation: Equipment Reliability 71152S The inspectors performed a review of potential adverse trends in the licensees corrective action program with emphasis on equipment issues tracked by the operations department and maintenance backlog lists.
In 2023, and again in 2024, the site identified an equipment reliability trend. Specifically, the site identified that bridging and mitigation strategies were sometimes not being effectively used to reduce risk in the period before final corrective actions for equipment issues. For example, the site identified that water was submerging cables in some manholes, and the site had not implemented bridging and mitigation strategies to keep the cables dry.
While the site typically corrects issues promptly, the inspectors determined that the site sometimes experiences delays in final corrective actions for more difficult problems that would reduce or eliminate the need for bridging and mitigation strategies. Specifically, in 2009, the licensee identified cracks in the concrete of a significant percentage of the manholes that contain wetted offsite power cables, non-safety related. Engineering subsequently developed an approved repair method to seal the cracks to stop water intrusion and repair the structure. However, in 2022 and 2024 the licensee identified again that the cracks had not been repaired and re-initiated action. The licensee has implemented bridging and mitigation strategies such as manually de-watering and installing sump pumps. The inspectors noted that the licensee still does not have work planned to fix the cracks, but the open work is to re-inspect the manholes and write new condition reports for the cracks.
However, the inspectors noted that many of the cracks are already documented and could be planned for repair. The lack of current scheduling to fix the condition will extend the potential for cables to be submerged and add additional inspection work to a long-standing issue. The inspectors concluded that ineffective maintenance prioritization and planning has created work that has not resolved the issue since 2009 and extended the bridging and mitigation strategy.
The licensee documented the inspectors observation in Condition Report CR-ANO-C-2025-00068. No issues of more than minor significance were identified.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On October 11, 2024, the inspectors presented the occupational radiation safety inspection results to Dana Swenszkowski, Radiation Protection Manager, and other members of the licensee staff.
- On November 18, 2024, the inspectors presented the ISFSI Follow-up Inspection for HI-STORM FW Version E1 canister noncompliance and URI closeout inspection results to Doug Pehrson, Site Vice President, and other members of the licensee staff.
- On November 26, 2024, the inspectors presented the Unit 2 inservice inspection results to Mark Skartvedt, General Manager Plant Operation, and other members of the licensee staff.
- On January 16, 2025, the inspectors presented the integrated inspection results to Doug Pehrson, Site Vice President, and other members of the licensee staff.
THIRD PARTY REVIEWS Inspectors reviewed the Institute on Nuclear Power Operations reports that were issued during the inspection period.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
60855
Calculations
CALC-22-E-0009-
Thermal Evaluation of HI-STORM FM Version E1 During a
Flood Accident at ANO
60855
Corrective Action
Documents
CR-ANO-
1-2023-01423, C-2023-03873, C-2023-03879, C-2024-
267, HQN-2023-03197
60855
Engineering
Changes
ECO-5018-130
HI-STORM FW Version E1 Engineering Change Order
60855
Engineering
Evaluations
CFR 72.48-
1541
HI-STORM FW Version E1 Variant
60855
Engineering
Evaluations
CFR 72.48-
1600
EC 91669, ANO HI-STORM FW Analysis
60855
Engineering
Evaluations
LBDCR-2022-034
EC 91669, HI-STORM FW Licensing & Operations at ANO
Unit 1
Corrective Action
Documents
CR-ANO-
1-2021-00661, 1-2022-01893, 2-2023-02513, 2-2024-01984,
C-2023-01883
Procedures
OP-1104.039
Plant Heating and Cold Weather Operations
Procedures
OP-1307.037
Unit 1 Plant Freeze Protection Testing
Procedures
OP-2106.032
Unit Two Freeze Protection Guide
Work Orders
Drawings
M-2230
Piping and Instrument Diagram Reactor Coolant System
Drawings
M-232
Piping and Instrument Diagram Decay Heat Removal System
111
Procedures
OP-1015.002
Decay Heat Removal and LTOP System Control
Procedures
OP-1015.008
Unit 2 SDC Control
Procedures
OP-1104.001
Decay Heat Removal Operating Procedure
140
Procedures
OP-2103.011
Draining the Reactor Coolant System
Procedures
OP-2104.004
Shutdown Cooling System
Corrective Action
Documents
CR-ANO-
2-2021-02787
Fire Plans
Unit 2 Prefire Plans
Miscellaneous
2CNA021502
Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment
Regarding Transition to a Risk-informed, Performance-based
2/18/2015
Miscellaneous
Arkansas Nuclear One - Unit 1 and Unit 2 FHA
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
CR-ANO-
1-2024-00529, 2-2006-00390, 2-2010-00597, 2-2010-00834,
2-2011-01818, 2-2015-00259, 2-2016-02663, 2-2017-00234,
2-2020-03059, 2-2021-03595, 2-2022-01016, 2-2022-01326,
2-2022-01797, 2-2023-00296, 2-2023-00381, 2-2023-00406,
2-2023-00627, 2-2023-00628, 2-2023-00630, 2-2023-00632,
2-2023-00700, 2-2023-00739, 2-2023-00744, 2-2023-00811,
2-2023-00858, 2-2023-00908, 2-2023-01096, 2-2023-01116,
2-2023-01141, 2-2023-01323, 2-2023-01402, 2-2023-01647,
2-2023-01670, 2-2023-01697, 2-2023-01730, 2-2023-01957,
2-2023-01961, 2-2023-01979, 2-2023-02048, 2-2023-02050,
2-2023-02051, 2-2023-02140, 2-2023-02359, 2-2023-02464,
2-2023-02465, 2-2023-02468, 2-2023-02472, 2-2023-02608,
2-2023-02609, 2-2023-02610, 2-2024-00530, 2-2024-00540,
2-2024-01058, 2-2024-01080, 2-2024-01145, 2-2024-01166,
C-2020-00244, C-2024-00753, C-2024-01341
Drawings
Arkansas Power and Light Company Arkansas Nuclear One
Unit 2, Hanger Detail Main Steam R 9 Sheets 1-5
Engineering
Changes
Operability Input CR-ANO-2-2021-03479 Tmin for SS Tubing
2EBB1: 2MS-1030F; 2 ft 1030
Miscellaneous
Request for Alternative to 10 CFR 50.55a(g)(6)(ii)(D)
Examination Requirements Arkansas Nuclear One, Unit 2
2/09/2009
Miscellaneous
Relief Request ANO2-RR-24-001, Half-Nozzle Repair of
Reactor Vessel Closure Head Penetration #71
10/21/2024
Miscellaneous
2CNA072101
Blind Zone SE 2CNA072101 ML21168A056 Arkansas
Nuclear One, Unit 2 - Approval of Request for Alternative
From Certain Requirements of the American Society of
Mechanical Engineers Boiler and Pressure Vessel Code
Miscellaneous
2R30PO6ISI
Scope-DMW Highlighted ANO PRG Agenda Summary
Report
04/23/2024
Miscellaneous
SEP-ISI-ANO2-105
Program Section for ASME Section XI, Division 1 ANO-2
Inservice Inspection Program
Procedures
CEP-NDE-0121
Reactor Vessel Upper Head (RVUH) Nozzle Penetration
Non-Visual Examination Guidelines
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
CEP-NDE-0404
Manual Ultrasonic Examination of Ferritic Piping Welds
(ASME XI)
Procedures
CEP-NDE-0423
Manual Ultrasonic Examination of Austenitic Piping Welds
(ASME XI)
Procedures
CEP-NDE-0641
Liquid Penetrant Examination (PT) for ASME Section XI
Procedures
CEP-NDE-0731
Magnetic Particle Examination (MT) for ASME Section XI
Procedures
CEP-NDE-0901
VT-1 Examination
Procedures
CEP-NDE-0902
VT-2 Examination
Procedures
CEP-NDE-0903
VT-3 Examination
Procedures
CEP-NDE-0955
Visual Examination (VE) of Bare-Metal Surfaces
308
Procedures
Boric Acid Corrosion Control Program (BACCP)
Procedures
Entergy Nuclear Welding Program
Procedures
EN-Ll-102
Corrective Action Program
53, 54
Procedures
Operability Determination Process
4, 20
Procedures
OP-2104.029
Service Water System Operations
27
Procedures
OP-2305.034
Service Water Boundary Valve Leak Test
Procedures
SEP-BAC-ANO-
001
Arkansas Nuclear One (ANO) Boric Acid Corrosion Control
Program Inspection and Identification of Boric Acid Leaks for
ANO-1 and ANO-2 R3
Corrective Action
Documents
CR-ANO-
2-2024-02195, 2-2024-02202
Miscellaneous
SES-1-056
Miscellaneous
SES-2-055
Procedures
EN-OP-115-14
Reactivity Management
Procedures
PWR Reactivity Maneuver
Procedures
OP-1102.002
Plant Startup
2
Procedures
OP-1102.008
Approach to Criticality
Procedures
OP-2102.004
Power Operation
Procedures
OP-2102.016
Reactor Startup
Procedures
OP-2103.015
Reactivity Balance
Procedures
OP-2302.002
Initial Criticality Following Refueling
Procedures
OP-2302.021
Sequence for Low Power Physics Testing Following
Refueling
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
CR-ANO-
1-2024-00704, 1-2024-00857, 2-2023-00475, 2-2023-01931,
2-2023-02397, 2-2023-02578, 2-2023-02787, 2-2024-00337,
2-2024-00721, 2-2024-00883, 2-2024-01066, 2-2024-01412,
2-2024-01868, 2-2024-02221
Procedures
Procedures
Quality Control Inspection Program
Work Orders
WO 54187946, 54194159, 54201248
CALC-09-E-0008-
ANO-1 NFPA 805 Non-Power Operations Assessment
Calculations
CALC-11-E-0006-
ANO-2 Start-Up 2 Fast and Manual Transfer Capability
Corrective Action
Documents
CR-ANO-
C-2024-01975
Miscellaneous
Tagout 0382
Day Auto Transformer Outage
10/03/2024
Procedures
COPD-024
Risk Assessment Guidelines
Procedures
Protected Equipment Postings
Procedures
Risk Informed Completion Time
Procedures
OP-1015.002
Decay Heat Removal and LTOP System Control
Procedures
OP-1015.048
Shutdown Operations Protection Plan
Procedures
OP-1107.001
Electrical System Operations
136
Procedures
OP-2107.001
Electrical System Operations
140
Work Orders
WO 573376, 52278714
Corrective Action
Documents
CR-ANO-
2-2023-02352, 2-2023-02453, 2-2023-02457, 2-2024-01663,
2-2024-02295, 2-2024-02296, 2-2024-02385, 2-2024-02386,
2-2024-02495, 2-2024-02514, 2-2024-02527
Engineering
Changes
Intake Structure Below El. 354
Miscellaneous
2R30 (A, B, C) SW Bay Inspection Report
Miscellaneous
Mini-Signal Generator, FM P/N 16302997
01/31/2024
Miscellaneous
TD A480.0020
Installation Operation and Maintenance Manual for Armco
Sluice Gates Heavy Duty Series
Miscellaneous
ULD-2-SYS-01
ANO-2 Emergency Diesel Generator System
Miscellaneous
ULD-2-SYS-10
ANO-2 Service Water System
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
OP-2104.029
Service Water System Operations
28
Procedures
OP-2104.036
Emergency Diesel Generator Operations
105
Procedures
OP-2302.034
Power Ascension Testing Controlling Procedure
Procedures
OP-2305.019
Service Water Pumps Flow Test
Procedures
OP-2411.102
Unit 2 Sluice Gate and SW Bay Cleaning and Inspection
Work Orders
WO 53026945, 54041638, 54187946
Miscellaneous
ANO-2 Cycle 31-
Arkansas Nuclear One, Unit 2, Cycle 31 Core Operating
Limits Report
Procedures
OP-2102.001
Plant Pre-Heatup and Pre-Critical Checklist
100
Procedures
OP-2102.002
Plant Heatup
Procedures
OP-2502.001
Refueling Shuffle
Work Orders
Calculations
CALC-A-2CV-
AOV Setpoints for 2CV-0301, 2CV-0302, 2CV-0303, 2CV-
0305, 2CV-0306
Corrective Action
Documents
CR-ANO-
1-2024-02096, 2-2023-01031, 2-2023-01687, 2-2023-01705,
2-2024-00025, 2-2024-00814, 2-2024-01179, 2-2024-01840,
2-2024-01900, 2-2024-02148
Engineering
Changes
AOV Sizing and Setpoints for 2CV-1001
Engineering
Changes
Documentation of Baseline for ANO2 Reference Values for
IST Components
Engineering
Changes
Maintenance Manual for Velan Forged/Cast Pressure Seal
Valves
Miscellaneous
AOV Setpoint Control, Signature Analysis and Trending
Miscellaneous
SEP-ANO-2-IST-1
Miscellaneous
SEP-ANO-AOV-
001
ANO Air Operated Valve Program
Miscellaneous
TDH219 0020
Hale FP1500DJ-TCL Manual for All Components
Procedures
Alternate Low Pressure Emergency Feedwater
Procedures
OP-1106.007
Common Feedwater System
Procedures
OP-2102.001
Plant Pre-Heatup and Pre-Critical Checklist
100
Procedures
OP-2104.036
Emergency Diesel Generator Operations
105
Procedures
OP-2104.039
HPSI System Operation
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
OP-2302.046
CEA Drop Time Test
Procedures
OP-2305.001
Integrated Engineering Safeguards Test
Procedures
OP-2305.003
ESF Response Time Test
Procedures
OP-2305.005
Valve Stroke and Position Verification
Procedures
OP-2305.017
Procedures
OP-2307.009
Pressurizer Proportional Heater Checkout
Work Orders
WO 569760, 594721-05, 52990661, 53023736, 53034213,
53036759, 53038901, 54014236, 54031397, 54031399,
54039218, 54057183, 54079296, 54082152, 54161329,
54163303, 54174695, 54187946, 54201213
ALARA Plans
RWP Termination
and Post-Job
ALARA Review:
RWP 20232420
RWP Descriptions: 2R29 Scaffold Activities
06/22/2023
ALARA Plans
RWP Termination
and Post-Job
ALARA Review:
RWP 20232430
RWP Description: 2R29 Reactor Dissassembly/Reassembly
06/29/2023
ALARA Plans
RWP Termination
and Post-Job
ALARA Review:
RWP 20232500
RWP Description: 2R29 Replace 2P-32A Motor
06/26/2023
Corrective Action
Documents
CR-ANO-
1-2023-00690; 1-2023-00878; 1-2024-01417; 2-2023-01904;
2-2024-00465; C-2023-01022; C-2023-01792; C-2024-00315
Corrective Action
Documents
Resulting from
Inspection
CR-ANO-
2-2024-01823, 2-2024-01824, 2-2024-01850
Procedures
Control of Non-Fuel Materials
Procedures
Radiation Worker Expectations
Procedures
Access Control for Radiologically Controlled Areas
Procedures
Personnel Contamination Events
Procedures
Radiological Work Permits
Procedures
Radiological Survey Documentation
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
EN-RP-106-01
Radiological Survey Guidelines
Procedures
ALARA Program
Procedures
Response to Contaminated Spills/Leaks
Procedures
Radioactive Material Control
Procedures
Alpha Monitoring
Procedures
Job Coverage
Procedures
Special Monitoring Requirements
Radiation Work
Permits (RWPs)
242407
2R30 Decontamination Activities
Radiation Work
Permits (RWPs)
242433
2R30 Remove and Cut-up Incore Detectors
Radiation Work
Permits (RWPs)
242471
2R30 Inspections on Unit 2 Reactor Head
Self-Assessments LO-ALO-2023-
00042
Radiation Inspection - RP - Radiological Hazards and
Exposure Controls, and In-Plant Airborne Radioactivity
Control and Mitigation. IP 71124.01 IP 71124.03
04/10/2024
Corrective Action
Documents
CR-ANO-
2-2023-00521, 2-2023-01904, C-2023-01792, C-2023-03176
Drawings
STM 1-09
Figure 09.01: Reactor Building Ventilation and Purge
Diagram
Drawings
STM 1-48
Figure 48.22A: Breathing Air Systems and Purification Loops
Drawings
STM 2-09
Containment Cooling and Purge Systems
Drawings
STM 2-47-3
Figure 1: Control Room Ventilation Normal Flow Path
Miscellaneous
ML-0311 Qual
Matrix
SCBA Qualification Matrix
Miscellaneous
Ultra Elite APR
Respirator
Operations and Instructions
Procedures
Radiation Worker Expectations
Procedures
EN-RP-106-01
Radiological Survey Guidelines
Procedures
Air Sampling
Procedures
Personnel Monitoring
Procedures
Respiratory Protection Program
Procedures
Inspection and Maintenance of Respiratory Protection
03/28/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Equipment
Procedures
Inspection and Maintenance of Respiratory Protection
Equipment
2/05/2024
Procedures
Inspection and Maintenance of Respiratory Protection
Equipment
3/30/2023
Procedures
Inspection and Maintenance of Respiratory Protection
Equipment
04/17/2023
Procedures
Inspection and Maintenance of Respiratory Protection
Equipment
04/03/2024
Procedures
Inspection and Maintenance of Respiratory Protection
Equipment
05/17/2023
Procedures
Inspection and Maintenance of Respiratory Protection
Equipment
05/05/2024
Procedures
Inspection and Maintenance of Respiratory Protection
Equipment
03/21/2023
Procedures
9.5
Annual Respiratory Protection Equipment Inventory and
Inspection
11/28/2023
Procedures
Selection, Issue and Use of Respiratory Protection
Equipment
Procedures
Breathing Air
Procedures
OP 1104.033 093
Procedures
OP 2104.035
Ventilation System Operations
Procedures
Procedure
1104.035
Fuel Handling and Radwaste Ventilation
Radiation Work
Permits (RWPs)
232465
2R29 - RCP Seal Replacement for 2P-32A, 2P-32B, and
Radiation Work
Permits (RWPs)
242053
Unit-2 Spent Fuel Activities
Radiation Work
Permits (RWPs)
242433
2R30 Remove and Cut-up Incore Detectors
Radiation Work
Permits (RWPs)
242471
2R30 Inspections on Unit 2 Reactor Head
Self-Assessments LO-ALO-2023-
00042
Radiation Inspection - RP - Radiological Hazards and
Exposure Controls, and In-Plant Airborne Radioactivity
04/10/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Control and Mitigation. IP 71124.01 IP 71124.03
71124.03
Work Orders
OP-5120.417 VEF-
38B
Perform 18M Test of VEF-38-B; Obtain Charcoal Sample
07/06/2023
Work Orders
VSF-9 Perform
Flow Traverse
20.415
18M VSF-9; In-place Testing of the Unit 1 Control Room
Filtration System
2/07/2022
Work Orders
18-month Test of 2vSF-9
20.425 2VSF-9 Charcoal Sample and Flow 18m Test
10/10/2023
Corrective Action
Documents
CR-ANO-
2-2024-00337, 2-2024-00883, 2-2024-00985, 2-2024-01688,
2-2024-01748
Engineering
Changes
Evaluation to Support 2CV-1050-2 Normally Closed
Corrective Action
Documents
CR-ANO-
1-2009-01996, 1-2009-02017, 1-2014-01680, 1-2014-01849,
1-2015-00071, 1-2020-00795
Corrective Action
Documents
CR-ANO-
2-2024-01738, 2-2024-01759, 2-2024-01963, 2-2024-01964,
C-2024-01777