IR 05000313/2025011

Age-Related Degradation Inspection Report 05000313/2025011 and 05000368/2025011

ML25129A009
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/13/2025
From:	Nick Taylor NRC/RGN-IV/DORS/EB1
To:	Pehrson D Entergy Operations
References
IR 2025011
Download: ML25129A009 (1)

Text

May 13, 2025

SUBJECT:

ARKANSAS NUCLEAR ONE - AGE-RELATED DEGRADATION INSPECTION REPORT 05000313/2025011 AND 05000368/2025011

Dear Doug Pehrson:

On April 10, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Arkansas Nuclear One and 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.

One finding of very low safety significance (Green) is documented in this report. This finding 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.

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, Nicholas H. Taylor Deputy Director Division of Operating Reactor Safety Signed by Taylor, Nicholas on 05/13/25 Docket Nos. 05000313 and 05000368 License Nos. DPR-51 and NPF-6

Enclosure:

As stated

Inspection Report

Docket Numbers:

05000313 and 05000368

License Numbers:

DPR-51 and NPF-6

Report Numbers:

05000313/2025011 and 05000368/2025011

Enterprise Identifier:

I-2025-011-0015

Licensee:

Entergy Operations, Inc.

Facility:

Arkansas Nuclear One

Location:

Russellville, AR

Inspection Dates:

March 24, 2025 to April 10, 2025

Inspectors:

W. Cullum, Senior Reactor Inspector

R. Kumana, Senior Reactor Inspector

D. Stagner, Reactor Inspector

Approved By:

Nicholas H. Taylor

Deputy Director

Division of Operating Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (programs) 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 Design Parameters Provide for Operation of the Atmospheric Steam Dump Block Valve 2CV-1052 Following a Main Steam Line Break Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000368/2025011-01 Open/Closed None (NPP)71111.21N.

The inspectors identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, when the licensee failed to ensure that high energy line break calculations support statements in the Safety Analysis Report for Main Steam Line Breaks (MSLB) in the Auxiliary Building.

Additional Tracking Items

None.

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 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

71111.21N.04 - Age-Related Degradation Age-Related Degradation

(1) Unit 2 Steam Generator B Atmospheric Dump Control Valve 2CV-1052

(2) Unit 2 Emergency Diesel Generator Mechanical Systems 2K-4A

(3) Unit 1 Decay Heat Pump Bearing Heat Exchanger E-50A

(4) Unit 2 4160V Switchgear 2A-3

(5) Unit 2 Emergency Diesel Generator Electrical 2K-4A

(6) Unit 1 Battery Charger D-04A

(7) Units 1 and 2 Safety-related Buried Cables

(8) Unit 2 Containment Sump Suction Isolation Valve 2CV-5650

(9) Unit 2 Service Water Pump 2P-4C

(10) Unit 2 Service Water Cross-Tie Valve 2CV-1421-2

(11) Unit 1 Sluice Gate SG-1

(12) Unit 2 Sluice Gate 2CV-1470-1

(13) Unit 1 Emergency Feedwater Pump P-7B

INSPECTION RESULTS

Failure to Ensure Design Parameters Provide for Operation of the Atmospheric Steam Dump Block Valve 2CV-1052 Following a Main Steam Line Break Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000368/2025011-01 Open/Closed None (NPP)71111.21N.0 The inspectors identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, when the licensee failed to ensure that high energy line break calculations support statements in the Safety Analysis Report for Main Steam Line Breaks (MSLB) in the Auxiliary Building.

Description:

Inspectors reviewed documentation associated with the Unit 2 Atmospheric Steam Dump Block Valve (ADV) 2CV-1052. Section 15.1.14 of the ANO SAR for Unit 2 describes Major Secondary System Pipe Breaks With or Without Concurrent Loss of AC Power. This section describes the use of ADVs following a MSLB. The following statements are made in this section:

Plant cooldown may be operator controlled via the remotely operated atmospheric steam dump valves.

Manual action may be taken to actuate the atmospheric steam dump and isolation valves 30 minutes subsequent to initiation of the accident.

Calculation CALC-87-EQ-0003-01 provides the analysis of high energy line breaks in the auxiliary building. Part of the analysis describes a MSLB in Room 2155, main steam isolation valve room or penthouse. This room contains the main steam isolation valves (MSIVs), the inboard ADVs and the main steam safety valves. The analysis states that, The large size of a main steam line break and length of time of the blowdown, assumed to be 30 minutes before operator action will isolate the blowdown, will maintain a superheated atmosphere for nearby components." The maximum pressure expected in Room 2155 following a MSLB is 15.9 psia, or 1.2 psig.

Some of the walls in Room 2155 are made of sheet metal paneling. A separate calculation, CALC-86-E-006-01, provides the analysis for the maximum internal pressure the steel siding will withstand for Room 2155. The conclusion of this calculation is that the steel siding will distort at pressures above 108 pounds/ft^2, or 0.75 psig.

Inspectors performed a walkdown of Room 2155 and noted that at least 2 different types of fasteners were used to secure the sheet metal to the building. Calculation 86-E-006-01 only describes one type of fastener to secure the sheet metal. Inspectors also identified that conduit was located adjacent to the sheet metal. The conduit was mounted to the sheet metal in several locations and also mounted to structural steel in the building. The additional fasteners and conduit supports were not accounted for in CALC-86-E-006-01. The additional structures that support the sheet metal walls will serve as reinforcement. Therefore, the walls may require a higher internal pressure before they deform and relieve the pressure from a MSLB. If the walls do not relieve the pressure as intended, it would challenge the assumption that an operator may enter the room to operate the ADV 30 minutes subsequent to the initiation of the accident. The calculation does not take into account an unisolable break in which the metal siding may buckle to relieve pressure, however an elevated temperature in the room could remain and impede manual operation of the ADV. The high energy line break calculation also does not consider smaller breaks in which the room would not pressurize enough for the metal siding to buckle and relieve pressure. Again, this could pose a challenge to manual operation of the ADV.

Inspectors also noted that the environment in Room 2155 is expected to be harsh following a MSLB with maximum temperature expected to be 319.4F. Even if the metal walls deform to provide some pressure relief, it is not clear that the temperature environment would be suitable for manual operator action to open the ADV following an unisolable rupture. Although the motor operator for 2CV-1052 was procured as an environmentally qualified (EQ)component, ANO decided not to include it in the EQ program. The component is not marked as EQ on the Master Equipment List. The motor was also procured as a quality component, however the station downgraded the motor in EAR-88-0663 to augmented quality. Therefore, it is unclear whether the ADV will function when operated remotely following a MSLB.

Since the functionality of the ADV may be lost with remote operation and the local temperature environment would not support manual operation, inspectors determined that the functionality of the ADV would be lost following an unisolable steam line rupture in room 2155.

Corrective Actions: The licensee has entered the deficiency into the corrective action program.

Corrective Action References: CR 2-2025-00574, CR 2-2025-00565

Performance Assessment:

Performance Deficiency: The licensee failed to ensure the adequacy of design of the atmospheric steam dump block valve 2CV-1052 such that operation of the valve is possible following an unisolable Main Steam Line Break in Room 2155 as required by 10 CFR 50 Appendix B, Criterion III, Design Control.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control 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 structural calculation for the sheet metal siding in Room 2155 did not account for additional types of fasteners and sections of conduit secured directly to the sheet metal found during the walkdown. This additional supporting structure would serve to reinforce the steel siding walls which may not relieve pressure as expected following a main steam line break in room 2155. Additionally, the high energy line break calculation only considers a main steam line break which is isolable in room 2155 by manual operator action within 30 minutes. It does not take into account an unisolable break in which the metal siding may buckle to relieve pressure, however an elevated temperature in the room could remain and impeded manual operation of the ADV. The high energy line break calculation also does not consider smaller breaks in which the room would not pressurize enough for the metal siding to buckle and relieve pressure. Again, this could pose a challenge to manual operation of the ADV. Inspectors determined that these issues with the calculations would have a meaningful and substantive impact to the remaining margin needed to relieve pressure and temperature in room 2155 following an unisolable main steam line break.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.

Inspectors used Exhibit 2, "Mitigating Systems Screening Questions". Inspectors determined that the issue represented a loss of the 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. Therefore, the issue was further screened using a detailed risk evaluation. A regional senior reactor analyst performed a detailed risk evaluation which characterized the finding to be of very low safety significance (Green). For the evaluation, the analyst assumed valve 2CV-1052 to be nonfunctional, by setting basic event MSS-MOV-CC-1052 to TRUE, during main steam line break events outside of containment and compared that result to the result of events where the valve would remain functional during such events for the maximum one year exposure time. The increase in core damage frequency was estimated to be 1.63E-8/year. The analyst ran this condition on the Arkansas Nuclear One, Unit 2, SPAR model, version 8.82, ran on SAPHIRE, version 8.2.11.

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 calculation which supported relieving pressure from Room 2155 following a MSLB was performed in 1986. Therefore, the performance deficiency is outside of the nominal 3 year period.

Enforcement:

Violation: Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in section 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the requirement above, from 1986 until 2025, the licensee did not establish design control measures that provided for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, the licensee did not account for the as-built configuration of Room 2155 to establish whether or not it could relieve pressure following a Main Steam Line Break. Additionally, the high energy line break calculations do not account for the temperature conditions following an unisolable steam line break which is complicated by the fact that 2CV-1052 is not being maintained as an environmentally qualified component.

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 April 10, 2025, the inspectors presented the design basis assurance inspection (programs) inspection results to Doug Pehrson and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

CALC-91-E-0099-

ECP Peak Temperature and Inventory Loss Analysis

Summary

CALC-91-R-2013-

Service Water Performance Testing Methodology

Calculations

CALC-V-2CV-

1052-10

MOV Torque Switch Setpoints for 2CV-1052

Corrective Action

Documents

(CR-ANO-)

2-2012-03336, 2-2013-01780, 2-2017-05364, 2-2019-

01302, 2-2019-01317, 2-2019-01574, 2-2020-03644, 2-

20-03660, 2-2021-03306, 2-2023-01479, 2-2020-

2904,2-2021-01194, 2-2021-02136, 2-2021-02294 2-2021-

2906, 2-2022-00397, 2-2022-01062, 2-2024-01141, 2-

24-01142_1, 2-2025-00269_1, 2-2021-02937,2-2021-

2565, C-2020-02354, 1-2020-01690, 1-2020-01704, 1-

22-00243, 1-2022-00794, C-2020-03267, C-2021-00623,

C-2021-02441, C-2022-01594, C-2023-00671, 2022-03434,

23-00048_1, 2023-00922_1, 2023-02860_1, 2023-

03417_1, 2024-01151_1, 2024-01713_1, 2025-00008_1

Corrective Action

Documents

Resulting from

Inspection

(CR-ANO-)

2-2025-00469, 2-2025-00485, 2-2025-00486, 2-2025-

00563, 2-2025-00575, 2-2025-00639, C-2025-00648, 2-

25-00565, 2-2025-00574

Drawings

E-1_SH1_REV63

Station Single Line Diagram

DCP 86-2017

Atmospheric Dump Valve Controls

DCP-80-2127

Weld in Seal Ring

06/09/1980

DCP-91-2010

Main Steam Motor Operated Valve Modification

EC-43811

Unit 1 and Unit 2 Service Water Pump Packing Evaluation

in Response to CR-ANO-2-2012-3336

Engineering

Changes

EC-45207

Upgrade the U1 and U2 SW Pump Shaft Sleeves to a

Coated Shaft Sleeve to Reduce Wear and Increase

Reliability

71111.21N.04

Engineering

Evaluations

CALC-ANO2-SA-

06-00001

ANO-2 Mitigating System Performance Index Basis

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

CALC-ANOC-SE-

23-00001

Maintenance Rule 10CFR50.65(a)(3) Periodic Assessment

January 2022 to 30 June 2023

ER-ANO-2005-

0708-000

2P-4A/B/C Packing and Shaft Sleeve Upgrade

SEP-APJ-002

ANO Primary Containment Leakage Rate Testing (Appendix

J) Program

SEP-MOV-ANO-

001

ANO Motor Operated Valve (MOV) Program

TD-SITE

PROGRAM

CABLE HEALTH

UPDATE

1Q/2025

ANO Medium Voltage Underground Cable Testing Program

(per EN-DC-346_REV-033125_EXECUTIVE SUMMARY)

TDA480.0020

Installation, Operation and Maintenance Manual for ARMCO

Sluice Gates

TDF019.0020

Instruction Manual for Service Water Pumps

Miscellaneous

TDR378.0060

Instruction Manual for Rotork A Range Actuators

Syncropack 1400 Series, Syncropak 1600 Series,

Syncroset

1403171.010A

Insulation Resistance Testing

EN-DC-184

NRC Generic Letter 89-13 Service Water Program

OP-1412.083

Rotork Valves and Valvops Inspection and Lubrication

OP-2104.029

Service Water System Operations

29

OP-2305.034

Service Water Boundary Valve Leak Test

OP-2402.034

Breakdown and Repair of Unit 2 Service Water Pumps (2P-

4A, B, and C)

OP-2402.094

Maintenance of Tricentric Butterfly Valves

Procedures

OP-2411.111

Inspection and Cleaning of Unit 2 Service Water Forebays

Work Orders

44088, 61846, 62422, 62812, 66325, 215010, 276276,

393112, 487547, 488033, 525042, 526963, 556719,

2570, 563358, 570897, 573821, 574710, 589740,

594232, 50269225, 51803722, 52190716, 52285988,

2466256, 52471638, 52593620, 52648356, 52712027,

2739341, 52748656, 441893, 541488, 52960541,

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

29970012, 53004629, 53012505