IR 05000237/2025003

Integrated Inspection Report 05000237/2025003 and 05000249/2025003 and Exercise of Enforcement Discretion

ML25364A360
Person / Time
Site:	Dresden
Issue date:	12/31/2025
From:	Robert Ruiz NRC/RGN-III/DORS/RPB1
To:	Mudrick C Constellation Energy Generation, Constellation Nuclear
References
EAF-RIII-2025-0124
Download: ML25364A360 (0)

Text

SUBJECT:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - INTEGRATED INSPECTION REPORT 05000237/2025003 AND 05000249/2025003 AND EXERCISE OF ENFORCEMENT DISCRETION

Dear Christopher H. Mudrick:

On September 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Dresden Nuclear Power Station, Units 2 and 3. On December 17, 2025, the NRC inspectors discussed the results of this inspection with H. Patel, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Due to the temporary cessation of government operations, which commenced on October 01, 2025, the NRC began operating under its Office of Management and Budget-approved plan for operations during a lapse in appropriations. Consistent with that plan, the NRC operated at reduced staffing levels throughout the duration of the shutdown. However, the NRC continued to perform critical health and safety functions and make progress on other high-priority activities associated with the ADVANCE Act and Executive Order 14300.

On November 13, 2025, following the passage of a continuing resolution, the NRC resumed normal operations. However, due to the 43-day lapse in normal operations, the Office of Nuclear Reactor Regulation granted the regional offices an extension on the issuance of the calendar year 2025, third quarter integrated inspection reports normally issued by November 15, 2025, to December 31, 2025. The NRC will resume the routine cycle of issuing inspection reports on a quarterly basis beginning with the calendar year 2025, fourth quarter integrated inspection reports, which will be issued 45 days after the fourth quarter ends on December 31, 2025.

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.

Four findings of very-low safety significance (Green) are documented in this report. Four of these findings involved violations of NRC requirements. Two Severity Level IV violations without an associated finding are documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

December 31, 2025 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 III; the Director, Office of Enforcement; and the NRC Resident Inspector at Dresden Nuclear Power Station, Units 2 and 3.

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 III; and the NRC Resident Inspector at Dresden Nuclear Power Station, Units 2 and 3.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Robert. Ruiz, Chief Reactor Projects Branch 1 Division of Operating Reactor Safety Docket Nos. 05000237 and 05000249 License Nos. DPR-19 and DPR-25 Enclosure:

As stated cc w/ encl: Distribution via LISTSERV Signed by Ruiz, Robert on 12/31/25

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Dresden Nuclear Power Station, Units 2 and 3, 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 Perform an Adequate Maintenance Risk Assessment for Corrective Maintenance Activities Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237/2025003-01 Open/Closed

[H.14] -

Conservative Bias 71111.13 Inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.65(a)(4)when the licensee failed to assess and manage the increase in risk prior to performing corrective maintenance activities on the Unit 2 high-pressure coolant injection (HPCI) gland seal leak-off (GSLO) system. This resulted in the licensee not performing risk management actions as required per OP-DR-201-012-1001, Dresden On-Line Fire Risk Management.

Failure to Identify and Correct a Condition Adverse to Quality Resulted in Submerging the High-Pressure Coolant Injection Gland Seal Leak-Off Drain Pump Motor Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237,05000249/2025003-02 Open/Closed

[H.11] -

Challenge the Unknown 71111.15 A self-revealed Green finding and an associated non-cited violation of 10 CFR Part 50,

Appendix B, Criterion XVI, Corrective Actions, was identified when the licensee failed to identify and correct a condition adverse to quality. Specifically, the licensee failed to identify and correct water intrusion into the high-pressure coolant injection (HCPI) room sump that resulted in the failure of the safety-related HPCI gland seal leak-off (GSLO) drain pump motor.

Primary Containment Isolation Valve Found Out of its Required Position Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000237,05000249/2025003-03 Open/Closed

[P.3] -

Resolution 71152A Inspectors identified a Green finding and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to perform activities affecting quality in accordance with documented instructions, appropriate for the circumstances, associated with the replacement of the Unit 3 A low pressure coolant injection (LPCI) pump suction relief valve 3-1501-13A. Specifically, the work order did not contain steps to verify relief valve 3-1501-13A was installed in its correct position, which resulted in the primary containment isolation valve (PCIV) being left open in its mode of applicability.

Failure to Promptly Identify and Correct a Condition Adverse to Quality Associated with Safety-Related Valve 2-1501-3A Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237,05000249/2025003-04 Open/Closed

[H.12] - Avoid Complacency 71152A A self-revealed Green finding and associated Non-cited violation (NCV) of 10 CFR Part 50,

Appendix B, Criterion XVI, was identified when the licensee failed to promptly identify and correct a condition adverse to quality associated with containment cooling service water (CCSW) valve 2-1501-3A. Specifically, the licensee failed to identify and correct a crack in the weld between the valve yoke and the actuator mounting plate following a failure of the valve to operate on November 14, 2023. As a result, the valve subsequently failed to operate again due to the same degraded condition on April 30, 2024.

Failure to Identify a Condition Adverse to Quality Regarding Exceeding Technical Specification Required Action Completion Time Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000237/2025003-05 Open/Closed Not Applicable 71152A A self-revealed Severity Level IV, Non-Cited Violation (NVC) of 10 CFR 50.73, Licensee Event Report System, was identified when the licensee failed to submit a Licensee Event Report (LER) following the determination that a system, structure, or component was not operable for greater than what was allowed by the sites Technical Specifications (TSs).

Specifically, following a past operability review of a failure associated with safety-related valve 2-1501-3A, on April 30, 2024, the licensee discovered that the valve was inoperable for greater than the 7 days allowed by Technical Specification (TS) LCO 3.7.1, Condition C.

However, the licensee failed to submit an LER within 60 days of discovering a condition reportable under 10 CFR 50.73.

Failure to Maintain Complete and Accurate Information Regarding Activities Affecting Quality Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000237,05000249/2025003-06 Open/Closed Not Applicable 71152A The inspectors identified a Severity Level IV, Non-Cited Violation (NVC) of 10 CFR 50.9,

Completeness and Accuracy of Information, for the failure of the licensee to maintain complete and accurate information related to safety-related valve 2-1501-3A. Specifically, following the replacement of the potentiometer on 2-1501-3A, on November 14, 2023, the licensee failed to generate an issue report identifying that the as-left condition of the potentiometer was outside of the prescribed band per the work instructions, which was a condition adverse to quality. Additionally, the licensee failed to retain the applicable portions of the work order detailing how the potentiometer was set up following replacement, which was an activity affecting quality.

Additional Tracking Items

Type Issue Number Title Report Section Status EDG EAF-RIII-2025-0124 Enforcement Action EAF-RIII-2025-0124:

Failure to Comply with 10 CFR Part 37 71124.08 Closed LER 05000249/2024-002-00 LER 2024-002-00 for Dresden Nuclear Power Station, Unit 3, Low Pressure Coolant Injection Pump Suction Relief Valve Inoperable as a Primary Containment Isolated Valve Resulted in a Condition Prohibited by Technical Specifications 71152A Closed LER 05000249/2024-002-01 LER 2024-002-01 for Dresden Nuclear Power Station, Unit 3, Low Pressure Coolant Injection Pump Suction Relief Valve Inoperable as a Primary Containment Isolation Valve Resulted in a Condition Prohibited by Technical Specifications 71152A Closed LER 05000249/2024-002-02 LER 2024-002-02 for Dresden Nuclear Power Station, Unit 3, Low Pressure Coolant Injection Pump Suction Relief Valve Inoperable as a Primary Containment Isolation Valve Resulted in a Condition Prohibited by Technical Specifications 71152A Closed LER 05000237/2024-001-01 LER 2024-001-01 for Dresden Nuclear Power Station, Unit 2, Containment Cooling Service Water Valve Failure Resulted in a Condition Prohibited by Technical Specifications and Loss of Safety Function 71153 Closed

LER 05000237/2024-001-00 LER 2024-001-00 for Dresden Nuclear Power Station, Unit 2, Containment Cooling Service Water Valve Failure Resulted in a Condition Prohibited by Technical Specifications and Loss of Safety Function 71153 Closed

PLANT STATUS

Unit 2 The Unit began the inspection period at full-rated thermal power where it remained with the exception of short-term power reductions for control rod sequence exchanges, testing, and as requested by the transmission system operator.

Unit 3 The Unit began the inspection period at full-rated thermal power. On September 4, 2025, the Unit was shutdown for maintenance outage D3M24 to replace degraded reactor recirculation pump seals. The Unit returned to full power on September 11, 2025, where it remained with the exception of short-term power reductions for control rod sequence exchanges, testing, and as requested by the transmission system operator.

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.

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 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) Fire zone (FZ) 9.0A, Unit 2 diesel generator elev. 517' on July 16, 2025

(2) FZ 8.2.5A, Unit 2 track way area elevation 517' and FZ 8.2.5E, Unit 3 track way area elevation 517' on July 22 and July 23, 2025.

(3) FZ 11.2.2, Unit 2 southeast corner room elevation 476' on August 6, 2025

(4) FZ 8.2.1A, Unit 2 condensate pumps elevation 469' on August 6, 2025

(5) FZ 11.2.3, Unit 2 HPCI pump room elevation 476' on August 8, 2025

Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the onsite fire brigade training and performance during an Unannounced Fire Drill on August 21, 2025.

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 downpower and rod sequence exchange on July 18, 2025.

(2) The inspectors observed and evaluated licensed operator performance in the control room during Unit 3 shutdown for maintenance outage D3M24 on September 3, 2025.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)

(1) The inspectors observed and evaluated licensed crew performance in the main control room simulator on August 26, 2025.

(2) The inspectors observed and evaluated licensed crew performance in the main control room simulator on September 30, 2025.

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 high-pressure coolant injection gland seal leak-off motor failure on August 18, 2025

(2) Unit 3 feedwater regulating valve back up nitrogen supply system on Aug 1, 2025 (3)condensate system under AR 4868729, 2C cond pump trip on overcurrent, on August 12, 2025

(4) Unit 3 station blackout diesel following the failure of a fire system piping flow switch under AR 4863076, U3 SBO day tank room flow switch degraded, on August 21, 2025

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)both units online risk green, fire risk green during unplanned 2/3 emergency diesel generator maintenance window on July 24, 2025

(2) Unit 2 emergent risk due to submerged gland seal leak-off motor associated with high-pressure coolant injection on July 1, 2025 (3)work week risk on September 4, 2025 (4)work week risk on September 15, 2025

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (14 Samples)

The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:

(1) Unit 3 High-Pressure Coolant Injection following torquing of minimum flow valve closed to address torus level trend as documented in AR 4879066, U3 Torus Water Level Increased Trend, on July 8, 2025

(2) AR 4885052, 2/3 Emergency Diesel Generator Transfer Logic Power Available Light Out, as documented in AR 4885052 on July 25, 2025 (3)spurious alarms on the control room 903-7 panel as documented in AR 4886711 and

===4886525 on July 28, 2025

(4) AR 4781180, DRE SLR - SFP Coupon Surveillance Program, on July 28, 2025

(5) AR 4890581, CCSW Keepfill Pressure Step Change, on August 14, 2025

(6) AR 4891911, OOTR 04829549-57 requires re-test, Aug 27, 2025

(7) AR 4893108, D2 LPRMs with Failing Power supplies left in operate, on August 27, 2025

(8) AR 4891045, MCR temperature is elevated, on August 28, 2025

(9) AR 4890532, SBO Manhole 1 cover not installed properly, August 29, 2025

(10) AR 4893327, 480V breaker EQ PM for Bus 28, 29, 38, 39 Sure-Trip trip units, on September 3, 2025

(11) AR 4895447, 2B Medium Range MCR Level Indicator Failing Upscale on September 5, 2025

(12) AR 4896013, 3-2301-5 Breaker Tripped While Opening, on September 12, 2025

(13) AR 4898414, U2 CCSW Div 1 thru wall leak, on September 17, 2025

(14) AR 4857081, Unit 2 HPCI Gland Seal Leak Off Motor Failed Megger, on September 18, 2025

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01)===

(1) The inspectors evaluated Unit 3 maintenance outage activities from September 4-September 9, 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) (7 Samples)

(1)post-maintenance testing on the Unit 2 emergency diesel generator following protective relay replacement and testing under WO 05402332 on September 5, 2025 (2)post-maintenance testing on the Unit 3 1A main steam isolation valve air manifold following troubleshooting associated with AR 4895657 on September 8, 2025 (3)post-maintenance testing on Unit 3 control rod E-3 following troubleshooting associated with AR 4895380 on September 9, 2025 (4)post-maintenance testing following replacement of relay 3-1530-284 under WO 1638373 on September 10, 2025 (5)post-maintenance testing following diagnostic testing of motor operated valve 3-1501-38B under WO 1836115 on September 11, 2025 (6)post-maintenance testing following repair of the torus vent valve supply air stop valve under WO 5671225-01 on August 24, 2025 (7)post-maintenance activities following troubleshooting and corrective maintenance associated with AR 4896171, 3-5741-19PT-1, Drywell Temp Point Failed, on the drywell atmospheric monitoring system on September 11, 2025

Surveillance Testing (IP Section 03.01) (2 Samples)

(1) DOS 1400-09, Core Spray System IST Comprehensive/Preservice Pump Test Torus Available, on August 7, 2025

(2) DOS 1300-05, 2/3B Isolation Condenser Makeup Pump Capacity Test, on August 21, 2025

Inservice Testing (IST) (IP Section 03.01) (1 Sample)

(1) Unit 2 containment cooling service water operability test and inservice test per DOS 1500-17 on July 1, 2025

71114.06 - Drill Evaluation

Required Emergency Preparedness Drill (1 Sample)

(1) Dresden integrated drill on August 7,

71124.07 - Radiological Environmental Monitoring Program

Environmental Monitoring Equipment and Sampling (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated environmental monitoring equipment and observed collection of environmental samples.

Radiological Environmental Monitoring Program (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the implementation of the licensees radiological environmental monitoring program.

GPI Implementation (IP Section 03.03) (1 Sample)

(1) The inspectors evaluated the licensees implementation of the Groundwater Protection Initiative (GPI) program to identify incomplete or discontinued program elements.

71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling,

Storage, & Transportation

Radioactive Material Storage (IP Section 03.01)

The inspectors evaluated the licensees performance in controlling, labeling and securing the following radioactive materials:

(1) Interim Radwaste Storage Facility

(2) Unit 1 Sealand Yard

Radioactive Waste System Walkdown (IP Section 03.02) (2 Samples)

The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:

(1)resins associated with the Advanced Liquid Processing System (ALPS) system (2)bags and barrels in the radwaste storage bins

Waste Characterization and Classification (IP Section 03.03) (2 Samples)

The inspectors evaluated the following characterization and classification of radioactive waste:

(1) ALPS resin (2)condensate resin

Shipment Preparation (IP Section 03.04) (1 Sample)

(1) The inspectors observed the preparation of radioactive shipment DM-25-044.

Shipping Records (IP Section 03.05) (4 Samples)

The inspectors evaluated the following non-excepted radioactive material shipments through a record review:

(1) DW-24-024; July 11, 2024; Clive; ALPS resin

(2) DW-24-016; September 17, 2024; Clive; condensate liner

(3) DW-24-044; December 1, 2024; Bear Creek; torus filter

(4) DM-25-001; January 8, 2025; Quad Cities; Kr-85 source

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

MS08: Heat Removal Systems (IP Section 02.07)===

(1) Unit 2 (July 1, 2024 through June 30, 2025)

(2) Unit 3 (July 1, 2024 through June 30, 2025)

MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)

(1) Unit 2 (July 1, 2024 through June 30, 2025)

(2) Unit 3 (July 1, 2024 through June 30, 2025)

MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)

(1) Unit 2 (July 1, 2024 through June 30, 2025)

(2) Unit 3 (July 1, 2024 through June 30, 2025)

BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)

(1) Unit 2 (April 1, 2024 through June 30, 2025)

(2) Unit 3 (April 1, 2024 through June 30, 2025)

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)review of licensee evaluation and corrective actions associated with primary containment isolation valve found out of its required position as documented in AR 4818810 on September 5, 2025 (2)review of licensee evaluation and corrective actions associated with a crack in the yoke weld of MOV 2-1501-3A as documented in AR 4838481 on September 23, 2025

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 in identifying issues at a low threshold that might be indicative of a more significant safety issue.

71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 05000249/2024-002-00, LER 05000249/2024-002-01, and LER 05000249/2024-002-02, Low Pressure Coolant Injection Pump Suction Relief Valve Inoperable as a Primary Containment Isolation Valve Resulted in a Condition Prohibited by Technical Specifications, (ADAMS Accession Nos. ML25022A141, ML25162A028, and ML25296A468). The inspection conclusions associated with this LER and its supplements are documented in this report under Inspection Results Section 71152A. This LER and supplement are Closed.

(2) LER 05000237/2024-001-00 and LER 05000237/2024-001-01, Containment Cooling Service Water Valve Failure Resulted in a Condition Prohibited by Technical Specifications and Loss of Safety Function (ADAMS Accession Nos. ML25027A435 and ML25296A231). The NRC evaluated this event and documented its initial assessment in an MD 8.3 Decision Documentation for Reactive Inspection (ML25063A119), dated February 27, 2025. The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71152A. This LER and supplement are Closed.

INSPECTION RESULTS

Failure to Perform an Adequate Maintenance Risk Assessment for Corrective Maintenance Activities Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237/2025003-01 Open/Closed

[H.14] -

Conservative Bias 71111.13 Inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.65(a)(4)when the licensee failed to assess and manage the increase in risk prior to performing corrective maintenance activities on the Unit 2 high-pressure coolant injection (HPCI) gland seal leak-off (GSLO) system. This resulted in the licensee not performing risk management actions as required per OP-DR-201-012-1001, Dresden On-Line Fire Risk Management.

Description:

On March 27, 2025, following a planned Unit 2 HPCI maintenance window, oil was noted to be leaking out of the HPCI turbine front standard console. The licensee identified that this oil came from a lube oil fitting worked during the maintenance window and tightened the fitting before restoring the system to service. As a result of the loose fitting, approximately 10 gallons of oil leaked out of the system and into the HPCI room sump, located under the front standard. The licensee removed power to the HPCI room sump pump to prevent inadvertently introducing oil into the radwaste system and implemented shiftly sump level checks, as directed by site procedure.

On April 2, 2025, the site received heavy rains which began to fill the HPCI room sump due to the sump being the collection point of the connecting reactor building roof drains. On April 5, 2025, the GSLO drain pump motor, located in the HPCI room sump pit, became partially submerged. The licensee wrote AR 4854009, U2 HPCI GSLO Drain Pump Wetted Due to Sump High Level, documenting the condition on April 7, 2025, and began remediating the condition by manually pumping down the water and oil mixture to drums. On April 14, 2025, the licensee tested the partially submerged GSLO drain pump motor. As documented in AR 4857081, U2 HPCI GSLO Motor Failed Megger, water and oil intrusion compromised winding resistance and resulted in the pump being unable to perform its design function.

The GSLO drain pump is safety related. The function of the GSLO drain pump is to return condensate from the GSLO condenser to the HPCI booster pump suction automatically based on condenser level when the HPCI system is in operation. The function of the GSLO condenser is to collect and condense steam from the HPCI turbine shaft seals, control valve stem leak-off, stop valve below seat drains, HPCI turbine casing drains, and turbine exhaust piping. In addition, the GSLO condenser works in conjunction with the GSLO exhauster to remove non-condensable gases from the condenser to the Standby Gas Treatment system, thus allowing the condenser to maintain a vacuum condition. The failure of the drain pump to perform its function would result in loss of GSLO condenser vacuum, leakage of steam into the HPCI room, and eventually a HPCI turbine trip on high room temperatures.

The licensee sent the GSLO drain pump motor offsite for refurbishment and restored it to service on April 30, 2025. The licensee declared the HPCI system operable and available without the GSLO drain pump based on engineering change (EC) 644198, Rev. 1, Loss of HPCI Gland Seal System and Impact Upon UFSAR Events.

On April 21, 2025, the resident inspectors reviewed EC 644198. The licensee conservatively determined that the time HPCI could operate without the GSLO drain pump in service before elevated temperatures in the room cause the turbine to trip was 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The licensee compared this run time to that assumed in UFSAR accidents that credit HPCI operation for post-accident response. Based on this comparison, the licensee determined that HPCI could meet its UFSAR-described functions without the GSLO drain pump in service and thus determined HPCI could be declared operable and available during the time the GSLO pump was inoperable.

Availability is defined via guidelines included in Attachment 5 of licensee procedure WC-AA-101, On-line Work Control Process. Attachment 5 references NUMARC 93-01, Section 11 for the purpose of guiding 10 CFR 50.65(a)(4) maintenance rule evaluations.

NUMARC 93-01 emphasizes the use of probabilistic safety assessments or probabilistic risk assessment (PRA) insights in 10 CFR 50.65(a)(4) evaluations, and licensee procedure WC-AA-101-1006, On-line Risk Management and Assessment, directs the use of PRA tools in 10 CFR 50.65(a)(4) evaluations at Constellation sites. As such, key safety functions defined by the licensee PRAs should be considered in availability determinations. Further, Inspection Manual Chapter (IMC) 0308, Attachment 3, Appendix K, Technical Basis for Maintenance Risk Assessment and Risk Management Significance Determination Process dated October 25, 2024, notes that a PRA [key safety] function refers to the ways in which the SSC can be used in a PRA to prevent an initiating event from resulting in core damage.

An SSC may have more or different PRA functions than those functions for which it is credited in the design or licensing basis. This acknowledges that evaluation for the purpose of maintenance risk may be too narrow if it focuses solely on functions credited in the design or licensing basis.

With these considerations in mind, the resident inspectors evaluated site-specific PRA data and EC 644198, Rev. 1 for the purpose of evaluating the licensees decision to declare the HPCI system available from April 7 through April 30, 2025, while maintenance on the Unit 2 HPCI GSLO drain pump motor was ongoing. DR-PSA-005.08, High Pressure Coolant Injection (HPCI) System Notebook, documents the basis for the HPCI system portion of the Dresden PRA. Section 6.0 specifically documents the basis for the HPCI unavailability model. It notes that the HPCI model will conservatively assume HPCI will isolate with a gland seal system failure... because temperatures in the room with an active steam leak would exceed the environmental qualification of controls in the room. Further, DR-PSA-010, Component Data Notebook, includes detail related to an environmental conditions result in HPCI failure fault tree. This tree includes logic to fail HPCI due, in part, to high room temperatures associated with GSLO failure. Finally, DR-PSA-002, Event Tree Notebook, documents the basis for licensee PRA event tree structure. Assumed system mission times related to the station black out (SBO) event tree are discussed in this notebook. Of note, certain [SBO] accident sequences use HPCI for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> plus an alternate mitigator to achieve success based on room cooling calculations that suggest HPCI can operate for more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during an SBO event. DR-PSA-005.08, the HPCI system notebook, also references the systems 4-hour mission time for SBO scenarios.

The inspectors noted that EC 644198, Rev. 1 does not discuss treatment of the GSLO system in the site PRA calculations, nor does it address the ability of HPCI to operate for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without the GSLO drain pump in service. The resident inspectors asked the licensee for additional information to address these observations; to date, the licensee has not produced additional information to support how EC 644198, Rev. 1 specifically addresses PRA assumptions. As such, the resident inspectors determined the HPCI system was not able to fulfill all of its designated PRA functions with the GSLO drain pump out of service and therefore, should have been declared unavailable while maintenance on the pump was ongoing between April 7 and April 30, 2025.

The HPCI system is identified in the sites PRA model as a fire risk-significant component.

Licensee procedure OP-DR-201-012-1001, Dresden On-line Fire Risk Management, defines an unavailable fire risk-significant component (UFRIC) as an unavailable mitigating system component with functions that are important to minimize core damage fire risk from fire initiators. Per direction in OP-DR-201-012-1001, if a UFRIC is scheduled for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the licensee is to identify and apply risk management actions (RMA) to mitigate the fire risk or to maintain the function of other available fire-risk-important components when UFRIC components are unavailable. In this case, RMA checklists 4, 5, 12, and 14 would have been directed to be performed once per shift during the ~23 days of GSLO pump/HPCI unavailability. Therefore, the inspectors determined that the licensee failed to assess and manage the increase in risk prior to the conduct of maintenance activities on the Unit 2 HPCI GSLO motor, resulting in the failure to perform RMAs as required.

Corrective Actions: The licensee has not entered the NRC observations related to HPCI availability into the corrective action system, but did refurbish the wetted GLSO motor, restoring system availability. Attachment 5 of WC-AA-101, On-line Work Control Process, was updated in August 2025, to include consideration of PRA mission times in determinations of system availability. This is unrelated to the performance deficiency described here.

Corrective Action References: AR 4854009 and AR 4857081

Performance Assessment:

Performance Deficiency: The licensees failure between April 7 and April 30, 2025, to assess and manage the increase in risk prior to the conduct of maintenance activities associated with the Unit 2 HPCI GSLO motor was contrary to 10 CFR Part 50.65(a)(4) and was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the 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 GSLO motor and GSLO subsystem is required for the HPCI system to perform all of its designated PRA functions. As such, the Unit 2 HPCI system was unavailable from April 7 through April 30, 2025, while the GSLO motor was out of service for repairs.

Inspectors reviewed the examples provided in IMC 0612, Appendix E, Examples of Minor Issues and determined that the more-than-minor discussion under example 8.d was appropriate because the inadequate risk assessment resulted in missed RMAs that would have otherwise been assigned per plant procedures.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix K, Maintenance Risk Assessment and Risk Management SDP. Section 04.01.02 of this router allows for an NRC evaluation of risk or a detailed risk evaluation (DRE). A senior reactor analyst (SRA) performed a DRE to evaluate the risk significance associated with a HPCI pump failure to run for the period between April 5 and April 30, 2025. The SRA used SAPHIRE version 8.2.12 and the Dresden SPAR model version 8.83 to assess the significance of the finding for all hazards except for fire, which is not represented in the SPAR model. Results from the licensees probabilistic risk assessment model were reviewed and considered best-available information to assess the significance of the finding for fire and large early release frequency (LERF). The following assumptions and factors were considered in the quantification:

a.

As noted, the finding was assumed to result in a failure to run for the high-pressure coolant injection (HPCI) pump because removal of the GSLO drain pump would result in a steam release into the HPCI room beyond the capacity of the room cooling system, resulting in loss of function due to an adverse system interaction (e.g., steam supply isolation on high temperature, governor failure). This was considered a conservative assumption because some scenarios do not require HPCI to run for an extended period (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

b.

Using a t + repair approach the exposure time was determined to be 25 days.

c.

No credit for FLEX was given because the FLEX strategy relies on operation of the HPCI pump.

d.

The Dresden SPAR model version 8.83 was modified with the assistance of Idaho National Laboratory staff to address a quantification issue with the seismic event trees. This was not considered a key assumption in the analysis.

The resultant change in core damage frequency (CDF) from a HPCI pump run failure was estimated to be 3.0 E-7/year for internal events, as well as 4.0 E-7/year for fire from the licensee model, for a total of 7.0E-7/year. The resultant change in LERF was estimated to be 1.3 E-9/year for internal events and 2.0 E-8/year for fire, for a total of 2.1E-8/year. Seismic, High Wind, and Tornado hazards were evaluated, but were not significant contributors. The dominant core damage sequences for the finding were driven by turbine building fires which result in a transient, due to loss of feedwater with an associated loss of the isolation condenser and suppression pool cooling. Additionally, operator failure to vent the containment leads to containment failure and unavailability of low-pressure injection, ultimately resulting in core damage.

IMC 0609, Appendix K uses an incremental core damage probability (ICDP) to screen maintenance risk assessment inspection findings, which is an annualized fraction of the CDF given a plant configuration for a given duration. The resident inspectors reviewed the plant configuration for the time period during which HPCI was unavailable using recorded operator logs. They determined, based on this review, that the ICDP for this case is equal to change in CDF calculated via the DRE. Given that the CDF and LERF changes calculated via DRE are

<1E-6 and <1E-7, respectively, the issue screens to very-low safety significance (Green).

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee had information documenting that the HPCI system cannot fulfill certain designated PRA functions without a functional GSLO subsystem. When inspectors questioned the availability of HPCI during corrective repairs of the subsystem, the licensee maintained that the HPCI system was available, without technical justification.

Enforcement:

Violation: Title 10 CFR 50.65(a)(4) states that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventative maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to SSCs that a risk-informed evaluation process has shown to be significant to public health and safety.

Contrary to the above, from April 7, 2025, to April 30, 2025, the licensee failed, before performing maintenance activities, to assess and manage the increase in risk that resulted from proposed maintenance activities. Specifically, the licensee failed to adequately assess the availability of the Unit 2 HPCI system during corrective maintenance on the GSLO drain pump motor, resulting in the failure to perform RMAs as required by site procedure OP-DR-201-012-1001.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Identify and Correct a Condition Adverse to Quality Resulted in Submerging the High-Pressure Coolant Injection Gland Seal Leak-Off Drain Pump Motor Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237,05000249/2025003-02 Open/Closed

[H.11] -

Challenge the Unknown 71111.15 A self-revealed Green finding and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, was identified when the licensee failed to identify and correct a condition adverse to quality. Specifically, the licensee failed to identify and correct water intrusion into the high-pressure coolant injection (HCPI) room sump that resulted in the failure of the safety-related HPCI gland seal leak-off (GSLO) drain pump motor.

Description:

On March 27, 2025, following a planned Unit 2 HPCI maintenance window, oil was noted to be leaking out of the HPCI turbine front standard console. The licensee identified that this oil came from a lube oil fitting worked during the maintenance window, and tightened the fitting before restoring the system to service. As a result of the loose fitting, approximately 10 gallons of oil leaked out of the system and into the HPCI room sump, located under the front standard. The licensee secured power to the HPCI room sump pump to prevent oil being introduced to the radwaste system and implemented shiftly sump level checks, as directed by site procedure.

On April 2, 2025, the site received heavy rains which began to fill the HPCI room sump due to the sump being the collection point of the connecting reactor building roof drains. On April 5, 2025, the GSLO drain pump motor, located in the HPCI room sump pit, became partially submerged. On April 5, 2025, a non-licensed operator noted in a log entry that HPCI sump level was Unsat. No action at that time was taken due to continued concerns surrounding introduction of oil into the radwaste system. On April 7, 2025, the licensee identified that the GSLO drain pump motor, located in the HPCI room sump pit, was partially submerged, wrote AR 4854009, U2 HPCI GSLO Drain Pump Wetted Due to Sump High Level, documenting the condition, and began remediating the condition by manually pumping down the water and oil mixture to drums.

The GSLO drain pump is safety related. The function of the GSLO drain pump is to return condensate from the GSLO condenser to the HPCI booster pump suction automatically, based on condenser level when the HPCI system is in operation. The function of the GSLO condenser is to collect and condense steam from the HPCI turbine shaft seals, control valve stem leak-off, stop valve below seat drains, HPCI turbine casing drains, and turbine exhaust piping. In addition, the GSLO condenser works in conjunction with the GSLO exhauster to remove non-condensable gases from the condenser to the Standby Gas Treatment system thus allowing the condenser to maintain a vacuum condition. The failure of the drain pump to perform its function would result in loss of GSLO condenser vacuum, leakage of steam into the HPCI room, and eventually a HPCI turbine trip on high room temperatures.

Corrective Actions: On April 14, 2025, the licensee tested the partially submerged GSLO drain pump motor. As documented in AR 4857081, U2 HPCI GSLO Motor Failed Megger, water and oil intrusion compromised winding resistance and resulted in the GSLO pump being unable to perform its design function. The licensee refurbished the installed motor and returned it to an operable status on April 30, 2025.

Corrective Action References: AR 4854009 and AR 4857081

Performance Assessment:

Performance Deficiency: The licensees failure starting on April 5, 2025, to identify and correct that increasing levels in the Unit 2 HPCI room sump prior could render the GSLO drain pump motor inoperable was contrary to 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the 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 licensees failure to identify and correct increasing HPCI room sump level trends rendered the safety-related GSLO drain pump motor unable to perform its design function and challenged the availability, reliability, and capability of the Unit 2 HPCI system.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the finding in accordance with IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, Section A and answered Yes to question 2, Does the degraded condition represent a loss of the PRA function of a single train TS system (such as HPCI/HPCS) for greater than its TS allowed outage time? Therefore, the finding required a detailed risk evaluation.

A senior reactor analyst (SRA) performed a detailed risk evaluation which characterized the issue as having very-low safety significance (Green). The analyst used SAPHIRE version 8.2.12 and the Dresden SPAR model version 8.83 to assess the significance of the finding for all hazards except for fire, which is not represented in the SPAR model. Results from the licensees probabilistic risk assessment model were reviewed and considered best-available information to assess the significance of the finding for fire and large early release frequency (LERF). The following assumptions and factors were considered in the quantification:

1. The finding was assumed to result in a failure to run for the HPCI pump because

removal of the GSLO drain pump would result in a steam release into the HPCI room beyond the capacity of the room cooling system, resulting in loss of function due to an adverse system interaction (e.g., steam supply isolation on high temperature, governor failure). This was considered a conservative assumption because some scenarios do not require HPCI to run for an extended period (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

2. Using a t + repair approach the exposure time was determined to be 25 days,

assuming that the GSLO motor was rendered inoperable beginning on April 5, 2025, until its restoration on April 30, 2025.

3. No credit for FLEX was given because the FLEX strategy relies on operation of the

HPCI pump.

4. The Dresden SPAR model version 8.83 was modified with the assistance of Idaho

National Laboratory staff to address a quantification issue with the seismic event trees. This was not considered a key assumption in the analysis.

The resultant change in core damage frequency (CDF) was estimated to be 3.0 E-7/year for internal events, as well as 4.0 E-7/year for fire from the licensee model, for a total of 7.0E-7/year. Therefore, the finding was determined to be of very-low safety significance (Green) since the total change in CDF was <1E-6/year. The resultant change in LERF was estimated to be 1.3 E-9/year for internal events and 2.0 E-8/year for fire, for a total of 2.1E-8/year. Therefore, the finding was determined to be of very-low safety significance (Green) since the total change in LERF was <1E-7/year. Seismic, High Wind, and Tornado hazards were evaluated, but were not significant contributors. The dominant core damage sequences for the finding were driven by turbine building fires, which result in a transient due to loss of feedwater with an associated loss of the isolation condenser and suppression pool cooling. Additionally, operator failure to vent the containment leads to containment failure and unavailability of low-pressure injection, ultimately resulting in core damage.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, on April 5, 2025, the licensee should have documented the condition within the corrective action program and addressed rising levels in the Unit 2 HPCI room sump before the GSLO drain pump motor became submerged and rendered unable to perform its design function.

Enforcement:

Violation: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, measures be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.

Contrary to the above, from April 5, 2025, until April 7, 2025, the licensee failed to promptly identify and correct a condition adverse to quality. Specifically, the licensee failed to identify and correct increasing water level into the HPCI room sump prior to submerging the HPCI GSLO drain pump motor and rendering it unable to perform its design function.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Enforcement Discretion Enforcement Action EAF-RIII-2025-0124: Failure to Comply with 10 CFR Part 37 71124.08

Description:

On April 27, 2016, the licensee was issued NRC Inspection Report Nos. 05000237/2016405 and 05000249/2016405, which documented a licensee-identified violation of 10 CFR Part 37, Physical Protection of Category 1 and Category 2, Quantities of Radioactive Material at Facilities with a 10 CFR Part 73, Physical Protection Program. The violation met the criteria for Enforcement Discretion as described in Enforcement Guidance Memorandum (EGM) 14-001, Interim Guidance for Dispositioning 10 CFR Part 37 Violations with Respect to Large Components or Robust Structures Containing Category 1 or Category 2 Quantities of Material at Power Reactor Facilities Licensed Under 10 CFR Parts 50 and 52.

Subsequently, the inspectors re-evaluated this activity and found that while a violation of regulatory requirements continues to exist, the conditions remain within the criteria for Enforcement Discretion established within the Interim Enforcement Policy (IEP) discussed within NRCs Enforcement Policy, Section 9.3.

Corrective Actions: As specified in the NRCs Enforcement Policy, Section 9.3, the application of discretion is authorized until the underlying technical issue is dispositioned through rulemaking or other regulatory action.

Enforcement:

Violation: On April 27, 2016, a violation of 10 CFR Part 37 was documented in Dresden Inspection Report Nos. 05000237/2016405 and 05000249/2016405. The inspectors determined that the previously identified violation remains.

Basis for Discretion: This violation met all of the criteria for Enforcement Discretion as described in NRCs Enforcement Policy, Section 9.3, Enforcement Discretion for Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material (10 CFR Part 37).

Primary Containment Isolation Valve Found Out of its Required Position Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000237,05000249/2025003-03 Open/Closed

[P.3] -

Resolution 71152A Inspectors identified a Green finding and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to perform activities affecting quality in accordance with documented instructions, appropriate for the circumstances, associated with the replacement of the Unit 3 A low pressure coolant injection (LPCI) pump suction relief valve 3-1501-13A. Specifically, the work order did not contain steps to verify relief valve 3-1501-13A was installed in its correct position, which resulted in the primary containment isolation valve (PCIV) being left open in its mode of applicability.

Description:

On November 20, 2024, while Unit 3 was in Mode 1 and during performance of system alignment inspection, the NRC inspectors identified that the A LPCI pump suction relief valve (3-1501-13A) appeared to be in a different position when compared to similar valves. The inspectors communicated the apparent discrepancy to the control room at 2:07 p.m. CST and, following investigation by the licensee, the valve was placed in its required position at 4:50 p.m. the same day. The site entered Technical Specification (TS) 3.6.1.3, Primary Containment Isolation Valves, Condition C for a containment penetration with one PCIV inoperable. The licensee documented the issue in AR 4818810, NRC Walkdown Question.

The valve has two functions, the first of which is to protect the 3A LPCI pump volume from over pressurization when the pump suction is isolated from the torus. The second function is as a primary containment isolation valve while shut. The suction relief valve has an integrated manual operating handle that allows manual lifting of the valve. With the valve found with the handle in the raised position, it could not perform its primary containment isolation function and was therefore inoperable.

The inspectors reviewed applicable work performed on the LPCI system during the refueling outage that ended shortly before the system alignment inspection. The 3-1501-13A relief valve was replaced on November 3, 2024, in accordance with work order (WO) 4593282, TS IST LPCI relief valve replacement 3-1501-3A, while the unit was in Mode 5. Inspector review of WO 4593282 noted that post-maintenance testing of the relief valve replacement was an external leak check of the flanged connection where the valve connected to the LPCI system. The inspectors determined that the WO instructions and post-maintenance testing did not address the position of the relief valve handle to ensure the valve was placed in its required position prior to mode transition or system restoration. The inspectors did not identify any other relief valve handles in the incorrect position. Also, extent of condition performed by the licensee on other relief valves worked during the outage did not identify any additional issues.

The inspectors reviewed TS 3.6.1.3 and noted that the limiting condition for operability requires that each PCIV be operable in Modes 1, 2, and 3. The licensee transitioned to Mode 2 at 2:02 a.m. CST on November 15, 2024 with the PCIV not in its required position.

They did not correct the condition within the TS 3.6.1.3 Condition C required action time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, nor completed Condition E required actions to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and were therefore in a condition prohibited by Technical Specifications at 6:02 p.m. on November 16, 2024, until November 20, 2024, at 4:50 p.m.

On September 5, 2025, the inspectors reviewed the Corrective Action Program Evaluation (CAPE) generated by AR 4818810 and noted that licensee evaluation determined that the relief valve did not create any additional flow paths that would divert LPCI pump discharge flow away from the reactor pressure vessel or torus, and did not negatively affect the available pump suction head. Thus, the 3A LPCI pump remained operable and available to perform its specified safety function. Following their review, the inspectors did not identify an issue with the licensees conclusion on pump operability.

The CAPE also noted that a similar event occurred in 2019 associated with a containment cooling heat exchanger tube side relief valve. The CAPE from that event, documented in AR 4297084, DEOP 200-01 Entry for Low Torus Water Level, identified the apparent cause to be insufficient work instructions during relief valve replacement. The inspectors reviewed this CAPE, which identified corrective actions to address model work orders and plant procedures. Specifically, corrective action 4297084-18 required a contractor organization to update all work orders for replacement of relief valves with lifting devices. The inspectors determined that the corrective action was completed on March 31, 2020; however, it narrowly focused on five specific work orders associated with Unit 3 feedwater heater relief valve replacement. It also did not include a step to verify proper positioning of lifting devices following installation.

The inspectors noted that the corrective action was reopened as a result of AR 4542719, 2022 PI&R Self Assessment - Corrective Action Closure Improvements, on December 14, 2022, based on the identification that the corrective action closure note did not specify that all model work orders for replacement relief vales were updated. The updated corrective action assignment required that corrective actions were understood, implemented and documented for all model work orders. The inspectors determined that the corrective action to update model work orders was not implemented for WO 4593282.

Corrective Actions: The licensee closed the 3-1501-13A valve and documented the inspector observations in AR 4818810, NRC Walkdown Question.

Corrective Action References: AR 4818810, AR 4542719, AR 4297084

Performance Assessment:

Performance Deficiency: The licensees failure to perform activities affecting quality in accordance with documented instructions, appropriate for the circumstances, associated with the replacement of PCIV 3-1501-13A was contrary to 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and was a performance deficiency.

Specifically, WO 4593282 did not contain instructions to ensure 3A LPCI pump suction relief valve was installed in its required position.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, PCIV 3-1501-13A was left open and could not perform its primary containment isolation function.

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

Specifically, the finding screened to Green because the inspectors answered No to both questions in Section C of Exhibit 3 of IMC 0609. The finding did not create an actual open pathway in containment, nor did it involve hydrogen igniters.

Cross-Cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee identified in December 2022, that effective corrective actions had not been completed associated with model work orders being updated with required steps to ensure relief valves were installed in the correct and required position. WO 4593282 was not updated in accordance with corrective actions. As such, PCIV 3-1501-13A was installed with the manual handle in the incorrect position on November 3, 2024.

Enforcement:

Violation: Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures.

The licensee established WO 4593282, TS IST LPCI relief valve replacement 3-1501-3A, Revision 0, as the implementing procedure for installing relief valve 3-1501-13A, an activity affecting quality.

Contrary to the above, from November 3 to November 20, 2024, the licensee failed to have a procedure for the installation of relief valve 3-1501-13A, a safety-related component.

Specifically, WO 4593282 did not have a step to verify relief valve 3-1501-13A was installed in its correct position.

Additionally, on November 15, 2024, the licensee transitioned to Mode 2, mode of applicability for LCO 3.6.1.3, with the PCIV open and failed to implement TS require actions on November 16, 2024. Consequently, the licensee was in violation of LCO 3.6.1.3 until 3-1501-13A was closed on November 20, 2024.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Promptly Identify and Correct a Condition Adverse to Quality Associated with Safety-Related Valve 2-1501-3A Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237,05000249/2025003-04 Open/Closed

[H.12] - Avoid Complacency 71152A A self-revealed Green finding and associated Non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, was identified when the licensee failed to promptly identify and correct a condition adverse to quality associated with containment cooling service water (CCSW) valve 2-1501-3A. Specifically, the licensee failed to identify and correct a crack in the weld between the valve yoke and the actuator mounting plate following a failure of the valve to operate on November 14, 2023. As a result, the valve subsequently failed to operate again due to the same degraded condition on April 30, 2024.

Description:

On November 14, 2023, while performing a flow verification surveillance on the containment cooling service water (CCSW) system during the D2R28 refueling outage, valve 2-1501-3A failed to open on system demand as designed. The licensee documented the issue in the corrective action program under issue report (AR) 4717507 and attributed the failure to an over-ranged potentiometer. Ultimately, the licensee replaced the potentiometer under work order (WO) 5417420 and restored the valve to service. Subsequently, on April 30, 2024, 2-1501-3A failed to open on demand during a startup of the CCSW system. The licensee documented the issue under AR 4770771 and again attributed the failure to an over-ranged potentiometer. On February 18, 2025, during equipment walkdowns, the licensee identified a crack in the weld between the valve yoke and the actuator mounting plate of 2-1501-3A. The cracked weld allowed the valve actuator to flex away from the valve body during operation, effectively increasing the stroke length of the valve and causing an intermittent condition where the valve potentiometer could become over ranged during operation. The licensee performed a root cause evaluation under issue report (AR) 4938481 and determined that the degraded condition associated with the cracked weld on 2-1501-3A was the cause of the valve failing to operate on November 14, 2023, and again on April 30, 2024.

The CCSW system works in conjunction with low pressure coolant injection (LPCI) to provide a means of removing energy from containment by providing cooling water to the LPCI heat exchangers. Valve 2-1501-3A is the heat exchanger discharge isolation valve and provides a means to control CCSW to LPCI differential pressure and CCSW flow within allowable limits.

The valve is normally closed but opens with the start of its associated CCSW pumps and is positioned as demanded by a differential pressure controller. A position feedback potentiometer provides input to the valve controller, which consists of a disk with 360 degrees of travel. During normal operation, the valve completes its stroke before the indication of valve position on the potentiometer traverses the full 360 degrees around the potentiometer, with additional margin to prevent over ranging. Should the length of the stroke inadvertently increase, the indication of valve position can traverse greater than 360 degrees around the potentiometer, essentially moving through the zero position and beginning another rotation, before the physical valve stroke is complete. This causes the potentiometer to send the incorrect valve position to the differential pressure controller, preventing the valve from operating as designed.

On September 23, 2025, the inspectors reviewed the licensees root cause report and circumstances surrounding the issue associated with valve 2-1501-3A. Regarding the failure on November 14, 2023, the licensee documented the condition in AR 4717507, but the inspectors determined that the licensee incorrectly attributed the failure to maintenance performed earlier in the outage under WO 1082615, which was unrelated to the potentiometer. The potentiometer was replaced under WO 5417420, and the valve was returned to service, but the licensee failed to identify and correct the cracked weld on 2-1501-3A. Upon review of the root cause under AR 4938481, the inspectors learned that the potentiometer was set up out of the prescribed tolerances when replaced under WO 5417420. The as-left values reduced the margin to over ranging of the potentiometer in the future. No issue report was documented regarding the nonconformance, and WO 5417420 was not retained by the licensee for review by the inspectors. The failure to maintain complete and accurate information is dispositioned in Inspection Results Section 71152A of this inspection report.

Following the failure of 2-1501-3A on April 30, 2024, the inspectors determined that the troubleshooting activities performed by the licensee under AR 4770771 also failed to identify and correct the cracked weld. In addition, the inspectors reviewed the past operability determination performed by the licensee under AR 4770771, which determined that 2-1501-3A was last operable following a successful stroke on April 2, 2024. The inspectors determined that the licensee failed to recognize that the 28 days between the last successful valve stroke and the failure on April 30, 2024, exceeded the sites Technical Specification (TS) limiting condition for operation (LCO) outage time of 7 days per TS LCO 3.7.1, Condition C. Licensee procedure PI-AA-125, Corrective Action Program (CAP) Procedure, revision 8, section 4.1.2, states, in part, that if at any time any question of current or past operability arises, then initiate an issue report. The inspectors determined that contrary to PI-AA-125, the licensee failed to initiate an issue report documenting a condition prohibited by TS related to the failure on April 30, 2023. The failure on April 30, 2024, also required a licensee event report to be submitted to the NRC 60 days following the event, per 10 CFR 50.73. On November 25, 2024, during an unrelated review, the licensee identified that reporting requirements under 10 CFR 50.73 were not performed regarding the April 30, 2024, failure and documented the condition under AR 48119830. The licensee subsequently submitted LER 237/2024-001-00 on January 27, 2025. The failure to submit a required report is dispositioned in the Inspection Results Section 71152A of this report.

Ultimately, the inspectors determined that the licensee failed to promptly identify and correct the crack in the weld between the valve yoke and the actuator mounting plate of 2-1501-3A, a condition adverse to quality, following the failure on November 14, 2023. As a result, the degraded condition caused an additional failure on April 30, 2024.

Corrective Actions: Following discovery of the cracked weld, the licensee removed the valve from service and conducted repairs under WO 5626370. The licensee performed extent-of-condition walkdowns under AR 4838780 on 2-1501-3B, 3-1501-3A, and 3-1501-3B, and no additional degraded conditions were noted. The licensee performed a root cause evaluation under issue report (AR) 4938481.

Corrective Action References: AR 4868481, MOV 2-1501-3A Crack in Yoke Weld; AR 4770771, 2-1501-3A Failed to Open; AR 4717507, 2-1501-3A Did Not Open

Performance Assessment:

Performance Deficiency: The inspectors determined that the failure to promptly identify and correct a crack in the weld between the valve yoke and the actuator mounting plate of 2-1501-3A was a violation of 10 CFR Part 50, Appendix B, Criterion XVI, and thus a performance deficiency. Specifically, the licensee failed to identify and correct the degraded condition associated with the cracked weld following a failure of the valve to operate on November 14, 2023. As a result, the valve subsequently failed again on April 30, 2024.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance 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 failure to promptly identify and correct the degraded condition associated with a cracked weld on November 14, 2023, resulted in an additional failure on April 30, 2024, adversely affecting the availability, reliability and capability of the CCSW system.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors reviewed IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and answered Yes to question 3, therefore, a detailed risk evaluation was performed.

A senior reactor analyst (SRA) performed a detailed risk evaluation which characterized the issue as having very-low safety significance (Green). The analyst used SAPHIRE version 8.2.12 and the Dresden SPAR model version 8.82 to assess the significance of the finding for all hazards with the exception of fire, which is not represented in the SPAR model.

Results from the licensees probabilistic risk assessment model were reviewed, with assistance from HQ staff, and considered best-available information to assess the significance of the finding for fire and large early release frequency (LERF). The following assumptions and factors were considered in the quantification:

The finding caused the division 1 CCSW heat exchanger outlet valve, MOV 2-1501-3A, to be unavailable for a period of 33.8 days. The basis for this assumption was that the degraded condition occurred intermittently and was dependent on the previous valve stroke to cause the valve position potentiometer to overrange, indicating open while the MOV was physically shut, leading to MOV 2-1501-3A to not open on demand. Using a t + repair approach the exposure time was determined by combining two periods of unavailability:

1. MOV 2-1501-3A was unavailable for 28.5 days from April 2nd, 2024, at 2:24 p.m. CST,

when the valve was cycled for a post-maintenance test, to May 1st, 2024, at 3:01 a.m., when repairs were completed following its failure to open on April 30th, 2024.

2. Additionally, MOV 2-1501-3A was unavailable for 5.3 days from October 24th, 2023, at

9:26 p.m., when the valve was cycled for maintenance, to October 30th, 2023, at 5:26 a.m.

when Dresden Unit 2 shutdown for a refueling outage. The period of time between the shutdown and when the failure revealed itself on November 14th, 2023, was reviewed but was judged to not contribute to the overall significance of the issue because MOV 2-1501-3A does not support a key safety function while shutdown.

The Dresden SPAR model version 8.82 was modified to more closely reflect assumptions in the licensee PRA model. Specifically, the shutdown cooling success criteria was changed from three, to one train required, based on DR-PSA-005.17 Rev 9, Dresden shutdown cooling system notebook. Additionally, the human error probability for initiation of the shutdown cooling system, which was noted to contain field actions such as filling the system and installing pump fuses, was changed from 5E-4 to 3E-2 using SPAR-H.

The licensee modified their fire model to implement guidance in National Fire Protection Association Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants, frequently asked question (NFPA 805 FAQ) 14-0009 on treatment of well-sealed motor control center (MCC) electrical panels greater than 440 VAC. The purpose of NFPA FAQ 14-0009 is to incorporate test data that informs the fraction of fires that escape the MCC to damage targets and progress in severity. This was considered a key assumption in the evaluation.

The SRA noted that the fire modeling was potentially conservative in that a non-suppression probability of 0.5, based on a zone of influence of 1 foot above the fire initiator, was used.

However, the most significant targets, associated with division 2 of CCSW system, were greater than 10 feet horizontally from the initiator. Based on the geometry of the room, this would result in division 2 of CCSW being impacted less frequently if more detailed fire modeling were done.

The resultant change in core damage frequency (CDF) was estimated to be 1.2 E-7/year for internal events and 7.0 E-7/year for fire, for a total of 8.2E-7/year. Therefore, the finding was determined to be of very-low safety significance (Green) since the total change in CDF was

<1E-6/year. The resultant change in LERF was estimated to be 1.7 E-9/year for internal events and 1.2 E-8/year for fire, for a total of 1.3E-8/year. Therefore, the finding was determined to be of very-low safety significance (Green) since the total change in CDF was

<1E-7/year. The dominant core damage sequences for the finding were driven by MCC fires which resulted in a plant transient due to loss of feedwater with an associated dual unit loss of offsite power. Additionally, these fires led to loss of suppression pool cooling due to damage to division 2 CCSW and inability to vent the containment leading to containment failure and unavailability of associated core cooling system, ultimately resulting in core damage.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, the licensee did not consider that a latent condition existed within 2-1501-3A causing the valve potentiometer to become over ranged during normal operation. As a result, the licensee failed to address the cause and the extent of condition related to over ranging the potentiometer.

Rather, the licensee attributed the individual failures to the operation of the potentiometer itself.

Enforcement:

Violation: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, measures be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, and defective material and equipment are promptly identified and corrected.

TS 3.7.1, Containment Cooling Service Water (CCSW) System, Limiting Condition for Operation (LCO) 3.7.1, states that that two CCSW subsystems shall be operable in Modes 1, 2, and 3. LCO 3.7.1, Condition C, states, if one CCSW subsystem is inoperable for reasons other than Condition A, the licensee is required to restore the CCSW subsystem to an operable status in 7 days. Condition D states, in part, that if the required actions and associated completion time of Condition C is not met, be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Contrary to the above, from November 14, 2023, to February 18, 2025, the licensee failed to promptly identify and correct a condition adverse to quality. Specifically, on November 14, 2023, following a failure of safety-related valve 2-1501-3A, the licensee failed to identify and correct a crack in the weld between the valve yoke and the actuator mounting plate, which was a condition adverse to quality. Subsequently, the valve failed again on April 30, 2024.

Additionally, contrary to the above, from approximately April 10, 2024, to April 30, 2024, the licensee was in a condition prohibited by Technical Specifications. Specifically, valve 2-1501-3A was determined to be inoperable following its last successful stroke on April 2, 2024, resulting in one subsystem of CCSW being inoperable for greater than the TS LCO allowed outage time of TS LCO 3.7.1, Condition C, and subsequently Condition D.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Identify a Condition Adverse to Quality Regarding Exceeding Technical Specification Required Action Completion Time Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000237/2025003-05 Open/Closed Not Applicable 71152A A self-revealed Severity Level IV, Non-Cited Violation (NVC) of 10 CFR 50.73, Licensee Event Report System, was identified when the licensee failed to submit a Licensee Event Report (LER) following the determination that a system, structure, or component was not operable for greater than what was allowed by the sites Technical Specifications (TSs).

Specifically, following a past operability review of a failure associated with safety-related valve 2-1501-3A, on April 30, 2024, the licensee discovered that the valve was inoperable for greater than the 7 days allowed by Technical Specification (TS) LCO 3.7.1, Condition C.

However, the licensee failed to submit an LER within 60 days of discovering a condition reportable under 10 CFR 50.73.

Description:

On April 30, 2024, safety-related valve 2-1501-3A, failed to operate as designed during the startup of the containment cooling service water system (CCSW). The licensee documented the condition in the corrective action program under issue report (AR) AR 4770771. On approximately June 20, 2024, the licensee completed a past operability determination under AR 4770771 and determined that 2-1501-3A was last operable following a successful stroke on April 2, 2024.

On September 25, 2025, the inspectors reviewed the causal evaluation the licensee performed under AR 4770771 and determined that the licensee had discovered that 2-1501-3A was not operable for 28 days between the last successful valve stroke and the failure on April 30, 2024, but failed to recognize that this period exceeded the LCO-required action completion time of 7 days per TS LCO 3.7.1. Condition C. Licensee procedure PI-AA-125, Corrective Action Program (CAP) Procedure, revision 8, section 4.1.2, states, in part, that if at any time any question of current or past operability arises, then initiate an issue report. The inspectors noted that no issue report was generated following the past operability review to further disposition the condition.

On November 25, 2024, during an unrelated review, the licensee identified that the reporting requirements under 10 CFR 50.73 were not performed and documented the condition under AR 48119830.

Ultimately, the inspectors determined that the licensee had discovered a condition prohibited by the sites TS following the past operability review performed under AR 4770771, and that the licensee failed to identify that the reporting requirements of 10 CFR 50.73 were applicable. The licensee submitted Licensee Event Report (LER) 237/2024-001-00 on January 27, 2025, to report the condition.

Corrective Actions: The licensee submitted LER 237/2024-001-00 on January 27, 2025.

Corrective Action References: AR 4819830, Missed LER for 2-1501-3A; AR 4770771, 2-1501-3A failed to open

Performance Assessment:

The inspectors determined this violation was associated with a minor performance deficiency.

Specifically, the failure to submit a required report in accorance with 10 CFR 50.73 was a performance deficiency. The performance deficiency was not viewed as a precursor to a more significant event, did not have the potential to lead to a more significant safety concern, and did not adversely affect a cornerstone objective listed in Inspection Manual Chapter 0612, Appendix B. Therefore, the performance deficiency was dispositioned by the inspectors as minor.

Enforcement:

The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.

Severity: The inspectors determined that NRC Enforcement Policy, Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, example d.9, was the most applicable to the violation. Therefore, the violation was determined to be Severity Level IV.

Violation: 10 CFR 50.73 states, in part, that the holder of an operating license under this part or a combined licensee under part 52 of this chapter for a nuclear power plant shall submit an LER for any event of the type described in this paragraph within 60 days after the discovery of the event. 10 CFR 50.73(a)(2)(i)(B) states, in part, any operation or condition which was prohibited by the plants Technical Specifications (TSs). 10 CFR 50.73(a)(2)(v)(B) states, in part, any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. 10 CFR 50.73(a)(2)(v)(D)states, in part, any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

Contrary to the above, between April 30, 2024, and November 24, 2024, the licensee was in a condition prohibited by Technical Specifications. Specifically, following the failure of the valve to open on April 30, 2024, the licensee completed a past operability review and discovered that the valve was not operable for greater than the required action completion time of Technical Specification LCO 3.7.1, Condition C and Condition D, but failed to take further action to disposition the issue. Consequently, greater than 60 days elapsed from the time of discovery of a condition prohibited by Technical Specifications under 10 CFR 50.73(a)(2)(i)(B), a condition which could have prevented the fulfillment of a safety function under 10 CFR 50.73(a)(2)(v)(B) and 10 CFR 50.73(a)(2)(v)(D), and the licensee submitting the required LER.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Maintain Complete and Accurate Information Regarding Activities Affecting Quality Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000237,05000249/2025003-06 Open/Closed Not Applicable 71152A The inspectors identified a Severity Level IV, Non-Cited Violation (NVC) of 10 CFR 50.9, Completeness and Accuracy of Information, for the failure of the licensee to maintain complete and accurate information related to safety-related valve 2-1501-3A. Specifically, following the replacement of the potentiometer on 2-1501-3A, on November 14, 2023, the licensee failed to generate an issue report identifying that the as-left condition of the potentiometer was outside of the prescribed band per the work instructions, which was a condition adverse to quality. Additionally, the licensee failed to retain the applicable portions of the work order detailing how the potentiometer was set up following replacement, which was an activity affecting quality.

Description:

On November 14, 2023, while performing a flow verification surveillance on the containment cooling service water (CCSW) system during the D2R28 refueling outage, valve 2-1501-3A failed to open on system demand as designed. The licensee documented the issue in the Corrective Action Program under issue report (AR) 4717507 and attributed the failure to an over-ranged potentiometer. Ultimately, the licensee replaced the potentiometer under work order (WO) 5417420 and restored the valve to service. Subsequently, on April 30, 2024, 2-1501-3A failed to open on demand during a startup of the CCSW system. The licensee documented the issue under AR 4770771 and again attributed the failure to an over-ranged potentiometer. On February 18, 2025, during equipment walkdowns, the licensee identified a crack in the weld between the valve yoke and the actuator mounting plate of 2-1501-3A. The licensee performed a root cause evaluation under issue report (AR) 4938481 and determined that the degraded condition associated with the cracked weld on 2-1501-3A was the cause of the valve failing to operate on November 14, 2023, and again on April 30, 2024.

On September 23, 2025, the inspectors identified that the licensees root cause evaluation discussed conditions adverse to quality that were not captured in the sites Corrective Action Program. Specifically, during the replacement of the potentiometer under WO 5417420, the root cause stated that the as-left potentiometer settings were established outside of the prescribed band of the work instructions, accepted by site maintenance and engineering, and no issue report was generated. The inspectors noted the stated cause of the failure on November 14, 2023, was that the potentiometer was over ranged. Therefore, setting up the replacement potentiometer out of the prescribed band resulted in reducing the margin to over ranging the potentiometer in the future. Although the as-left condition of the potentiometer, by itself, did not result in the loss of function of 2-1501-3A, the inspectors considered the as-left set up of the potentiometer a nonconformance with the work instruction and a condition adverse to quality requiring an issue report in accordance with licensee procedure PI-AA-125, Corrective Action Program (CAP) Procedure, Revision 8. The inspectors also noted that the as-left condition of the potentiometer was not recorded in WO 5417420. Following inspector questioning, the licensee was unable to locate the applicable portions of WO 5417420 pertaining to how the potentiometer was set up after being replaced, which was an activity affecting quality.

The failure to document a condition adverse to quality in the Corrective Action Program and the failure to retain evidence of quality maintenance on a safety-related component resulted in incomplete information available to be reviewed by the inspectors at the time of the failures. Had the information been maintained completely and accurately by the licensee, it would have likely caused the NRC to increase the scope of inspection activities following the failure of 2-1501-3A on November 14, 2023, and again on April 30, 2024, to understand how the corrective actions taken by the site addressed the degraded condition.

Corrective Actions: Following an unsuccessful search for the missing sections of work order 54178420 documenting the potentiometer setup, the licensee dispositioned the applicable sections as lost under issue report (AR) 4902813.

Corrective Action References: AR 4902813, NRC ID: WO 5417420 Partial Lost Record; AR 4717507, 2-1501-3A Did Not Open; AR 48384891, MOV 2-1501-3A Crack in Yoke Weld

Performance Assessment:

The inspectors determined this violation was associated with a minor performance deficiency.

Specifically, the inspectors determined that the failure to maintain complete and accurate information as required by 10 CFR 50.9 was a performance deficiency. The performance deficiency was not viewed as a precursor to a more significant event, did not have the potential to lead to a more significant safety concern, and did not adversely affect a cornerstone objective listed in Inspection Manual Chapter 0612, Appendix B. Therefore, the performance deficiency was dispositioned by the inspectors as minor.

Enforcement:

The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. Specifically, had the information been maintained completely and accurately by the licensee, it would have likely caused the NRC to increase the scope of inspection activities following the failure of 2-1501-3A on November 14, 2023, and again on April 30, 2024.

Severity: The inspectors determined that NRC Enforcement Policy, Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, example d.1, was the most applicable to the violation in that had the licensee maintained complete and accurate information, the inspectors would have increased their scope of inspection associated with the licensees troubleshooting activities. Therefore, the violation was determined to be Severity Level IV.

Violation: Title 10 CFR 50.9 (a), Completeness and Accuracy of Information, requires information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commissions regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.

Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, stated, Activities affecting quality to 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.

The licensee established PI-AA-125, Corrective Action Program (CAP) Procedure, Revision 8, as the implementing procedure for the Corrective Action Program. PI-AA-125, Section 4.1.2, directs, in part, if at any time a condition adverse to quality arises, then initiate an issue report in accordance with PI-AA-120. Licensee procedure PI-AA-120, Issue Identification and Screening Process, Revision 13, is the implementing procedure for identifying, screening, and classification of conditions adverse to quality. PI-AA-120, Section 2.0, Terms and Conditions, Subsection 2.3, defines a condition adverse to quality, in part, as an all-inclusive term in reference to any of the following: failures, malfunctions, deficiencies, defective items, and nonconformances.

Title 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records, states, in part, sufficient records shall be maintained to furnish evidence of activities affecting quality.

Licensee procedure RM-AA-101, Records Management Program, Revision 11, is the implementing procedure to manage records providing evidence of quality generated by the licensee. RM-AA-101-1008, Processing and Storage of Records, Revision 9, Section 4.7.3, states that completed records shall be turned over to records management per RM-AA-101 requirements.

Contrary to the above, on approximately November 14, 2023, the licensee failed to maintain accurate information required to be maintained by the Commissions regulations in all material respects. Specifically, the licensee failed to initiate an issue report regarding the as-left potentiometer settings being set outside of the prescribed band in WO 5417420, which was a condition adverse to quality. Additionally, the licensee failed to maintain WO 5417420, a record providing evidence of quality activities, following the replacement of the potentiometer. This information was material to the NRC because it would have likely caused the NRC to increase the scope of inspection activities following the failure of 2-1501-3A on November 14, 2023, and again on April 30, 2024.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Minor Violation 71152S Minor Violation: The final NRC Safety Culture Policy Statement (SCPS) that was published on June 14, 2011, provides the NRCs expectation that individuals and organizations performing regulated activities establish and maintain a healthy safety culture that recognizes the safety and security significance of their activities and the nature and complexity of their organizations and functions. Because safety and security are the primary pillars of the NRCs regulatory mission, consideration of both safety and security issues, commensurate with their significance, is an underlying principle of the SCPS. NUREG-2165, Safety Culture Common Language, lists the traits, attributes, and examples of a healthy nuclear safety culture.

10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures.

Licensee procedure NO-AA-10, Quality Assurance Topical Report (QATR), Revision 98, Chapter 16, Corrective Action, Section 16.2.1, states, in part, that the company implements a Corrective Action Program to promptly identify and correct items or occurrences that are adverse to quality. Licensee procedure PI-AA-125, Corrective Action Program (CAP)

Procedure, is the implementing procedure for the licensees Corrective Action Program which prescribes activities affecting quality.

In accordance with Inspection Procedure (IP) 71152, Problem Identification and Resolution, Section 03.02, the inspectors performed a semiannual trend analysis of the licensees CAP to identify any potential trends that might indicate the existence of a more significant safety issue. During their review, inspectors determined that there was a negative trend in the area of PI.1, Identification, of NUREG-2165, where the organization implements a CAP with a low threshold for identifying issues. Specifically, individuals ensure that issues, problems, degraded conditions, and near-misses are promptly recorded and documented in the CAP at a low threshold. Specific examples reviewed by the inspectors are provided below:

1. Work Order (WO) 9912153, D2 29Y EQ LPCI 1501-3A VLV Potentiometer

Replacement, dated April 4, 2022, documented as-left resistance readings outside of the prescribed band on the potentiometer for safety-related valve 2-1501-3A, which was a condition adverse to quality, and did not initiate an issue report (AR) contrary to PI-AA-125, Corrective Action Program (CAP) Procedure, Revision 7.

2. WO 5417420, 2-1501-3A Failed to Open, dated November 14, 2023, documented

as-left resistance readings outside of the prescribed band on the potentiometer for safety-related valve 2-1501-3A, which was a condition adverse to quality, and did not initiate an issue report (AR) contrary to PI-AA-125, Corrective Action Program (CAP)

Procedure, Revision 8.

3. AR 4770771, 2-1501-3A Failed to Open, dated April 30, 2024, documented that

safety-related valve 2-1501-3A was inoperable for 28 days during a past operability review but failed to recognize that this period exceeded the limiting condition for operation (LCO) required action completion time of 7 days per Technical Specification (TS) LCO 3.7.1, Condition C, which was a condition adverse to quality. The licensee failed to initiate an issue report (AR), contrary to PI-AA-125, Corrective Action Program (CAP) Procedure, Revision 8.

4. WO 5550355, EMD Adjust MOV 2-1501-3A Limits/Stroke Length, dated

December 5, 2024, identified grade 3 stem lube on 2-1501-3A, a safety-related valve.

Licensee procedure MA-AA-723-300, Diagnostic Testing of Motor Operated Valves, Revision 14, directs technicians to initiate an issue report for stem lube grade 3 or higher. Contrary to MA-AA-723-300, an activity affecting quality, no issue report was initiated in accordance with PI-AA-125, Corrective Action Program (CAP) Procedure, Revision 9.

5. AR 4885552, Troubleshooting Action Not Timely, dated July 24, 2025, documented

an event during which an operator performing rounds discovered that the light bulb associated with the safety-related 2/3 emergency diesel generator transfer logic power indication appeared to be burnt out. Since a replacement bulb was not immediately available, the operator removed the bulb, confirmed it was burnt out, and then reinserted the bulb into the panel, upon which they immediately received several unexpected alarms. The operator reset the alarms, failed to review the associated alarm response procedures, and incorrectly marked in the shift rounds that the burnt-out light indicated SAT. The light bulb remained extinguished for the remainder of the shift and was not turned over to the oncoming crew. During the next shift, a different operator identified the burnt-out light bulb on rounds, replaced it with a new bulb, but was unsuccessful in getting the light to illuminate. The operator notified the control room upon which troubleshooting revealed that a blown fuse had occurred in the panel. The blown fuse, a condition adverse to quality, was documented in the Corrective Action Program under AR 4885052 an entire shift after it had first occurred.

PI-AA-125, Corrective Action Program (CAP) Procedure, Revision 9, Section 4.1.2, directs, in part, if at any time a condition adverse to quality arises, then initiate an issue report in accordance with PI-AA-120. Licensee procedure PI-AA-120, Issue Identification and Screening Process, Revision 13, Step 4.3.3, directs individuals to contact their immediate supervisor to ensure necessary immediate actions are taken and appropriate routing is applied. Step 4.3.4 directs individuals to originate an issue report. Step 4.3.6 directs individuals to notify the shift and route the issue for immediate review by Operations Shift Management if immediate actions are required by Operations. The inspectors determined that an issue report for the condition adverse to quality was not documented as required by PI-AA-125 and not routed and reviewed per PI-AA-120, upon receipt of unexpected indications requiring immediate Operations actions following the insertion of the burnt-out light bulb back-in the panel.

As a result, the condition remained unevaluated by Operations until discovered and investigated by the next crew.

The inspectors concluded that the above examples demonstrated a failure to document issue reports as prescribed by instructions and procedures, were contrary to 10 CFR Part 50, Appendix B, Criterion V, and constituted a performance deficiency.

Screening: The inspectors determined the performance deficiency was minor. Specifically, the failure to document issue reports as prescribed by instructions and procedures was not viewed as a precursor to a significant event, did not have the potential to lead to a more significant safety issue, and did not adversely affect a cornerstone objective.

Enforcement:

The examples identified by the inspectors were discussed with the licensee and are either historic in nature, have been superseded by additional casual evaluations performed by the licensee in accordance with the Corrective Action Program, or are minor in significance. Therefore, the failure to comply with 10 CFR Part 50, Appendix B, Criterion V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On September 24, 2025, the inspectors presented the radiation protection inspection results to H. Patel, Site Vice President, and other members of the licensee staff.

• On December 17, 2025, the inspectors presented the integrated inspection results to H. Patel, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Corrective Action

Documents

AR 4894735

4.0 Critique Crew 4 Q3 Fire Drill

09/02/2025

Corrective Action

Documents

Resulting from

Inspection

AR 4885251

NRC ID: Fire Seal F-138-3 Degraded

07/24/2025

103 U2RB-3

Unit 2 Southeast Corner Room Elev. 476'

104 U2RB-4

Unit 2 HPCI Pump Room Elev. 476'

133 U2TB-36

Unit 2 Condensate Pumps Elev. 469'

140 U2TB-43

Dresden Pre-Fire Plan Layout Unit 2 Diesel Generator

Elev. 517'

2 U2TB-45

Unit 2 Track Way Area Elev. 517'

Fire Plans

164 U3TB-75

Unit 3 Track Way Area Elev. 517'

Miscellaneous

Fire Drill Scenario: U3 SWGR-35 Cubicle 6A 5A

07/17/2025

WO 5107155-01

EM D2 3Y TSTR/Com Emer D/G CARDOX System

Maintenance Test

2/05/2023

71111.05

Work Orders

WO 5312615

D2/3 2Y TSTR U2, 3, &2/3 D/G Room and U2&U3 Alterex

Heat Detec

04/12/2024

CPE OBE 25-05A

Dresden Operations Training LORT 2025 CPE OBE A

D229-022

ReMA D2C29 EOC Rod Pull #1

07/17/2025

Miscellaneous

D229-022

Control Rod Sequence Rod Move Sheet

07/15/2025

DGP 02-01

Unit Shutdown

179

DGP 02-03

Reactor SCRAM

21

DGP 03-01

Power Changes

141

DOP 1000-03

Shutdown Cooling Mode of Operation

71111.11Q

Procedures

DOP 4400-08

Circulating Water System Flow Reversal

AR 04863076

U3 SBO Day Tank Room Flow Switch Degraded

05/05/2025

AR 4799845

U3 FWRV Back-up N2 Supply PRV Failing

09/06/2024

AR 4821445

FW REG VLVS Backup N2 Supply PRV 3-4799-1748 is

Failing

01/23/2024

71111.12

Corrective Action

Documents

AR 4835830

FW REG VLVS Backup N2 Supply PRV 3-4799-1748 is

Failing

2/08/2025

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

AR 4854009

U2 HPCI GSLO Drain Pump Wetted due to Sump High Level

04/07/2025

AR 4857081

U2 HPCI GSLO Motor Failed Megger

04/14/2025

AR 4865012

N2 Supply PRV 3-4799-1748 Failure Analysis

2/08/2025

AR 4868729

2C Cond Pump Trip on Overcurrent

DR-27D-M-002

Dresden Station Blackout (SBO) Building Ventilation Air

Requirement

AR 4890290

PRV 3-4799-1748 Failure Analysis

08/13/2025

AR 4890705

NRC ID: Missed HSS-CME on Maint Rule Screening of

IR 4799845

08/14/2025

Corrective Action

Documents

Resulting from

Inspection

AR 4900688

NRC ID: U3 SBO MRULE Screening

09/23/2025

Engineering

Evaluations

EC 644198

Loss of HPCI Gland Seal System and Impact Upon UFSAR

Events

ER-AA-320

Maintenance Rule Implementation Per NEI 18-10

Procedures

ER-AA-320-1001

Maintenance Rule 18-10 - Scoping

AR 4885052

2/3 EDG Transfer Logic Power Available Light Out

07/24/2025

Corrective Action

Documents

AR 4898414

U2 CCSW Div 1 Thru Wall Leak

09/17/2025

Engineering

Changes

EC 396149

Technical Evaluation on Control of Temporary Openings in

Secondary Containment

DAP 07-44

Control Of Temporary Openings in Secondary Containment

During Performance of Work Packages, Surveillances, Or

Other Procedures

OP-AA-101-111

Roles and Responsibilities of On-Shift Personnel

OP-AA-108-117

Protected Equipment Program

OP-DR-201-012-

1001

Dresden On-line Fire Risk Management

71111.13

Procedures

WC-AA-101

On-Line Work Control Process

Calculations

DRE-3-2301-14

Handwheel Torque Calculation DC Motor Operated GL96-05

Globe Valve with SMB-0

007

AR 00954138

Procedure Deficiency-Manhole Inspection/Dewatering

07/07/2009

AR 01081708

Submerged Cables Found in Manholes

06/17/2010

AR 4781180

DRE SLR - SFP Coupon Surveillance Program

06/17/2024

AR 4818676

Trend IR: SBO-MH1 Water Level

11/20/2024

71111.15

Corrective Action

Documents

AR 4845269

U3 CCSW Keepfill Pressure Low After ECCS Clr Flow

03/14/2025

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Balancing

AR 4857081

U2 HPCI GLSO Motor Failed Megger

04/14/2025

AR 4862864

2B Medium Range MCR level Indicator Failing Upscale

05/03/2025

AR 4869351

WO to Adjust 3-3999-251A to Improve U3 CCSW Keepfill

Press

05/29/2025

AR 4879066

U3 Torus Water Level Increased Trend

07/04/2025

AR 4885052

2/3 EDG Transfer Logic Power Available Light Out

07/24/2025

AR 4886525

Spurious Alarm 903-7 D-1 HPCI Turb Supv Trouble

07/28/2025

AR 4886711

Spurious Alarm 903-7 C-1 U3 Diesel Gen Clg Wtr PP

Trip/Lkout

07/29/2025

AR 4890532

SBO Manhole 1 Cover Not Installed Properly

08/11/2025

AR 4890581

CCSW Keepfill Pressure Step Change

08/07/2025

AR 4891045

MCR Temperature is Elevated

08/17/2025

AR 4891360

Unit 2 LPRM 40-25A ICPS Found Reading Only 17 Vdc

08/19/2025

AR 4891781

Unit 2 LPRM 48-17D ICPS Found Reading Only 61 Vdc

08/20/2025

AR 4891911

OOTR 04829549-57 Requires Re-Test

08/20/2025

AR 4893108

D2 LPRMs with Failing Power Supplies Left in Operate

08/25/2025

AR 4893327

480V Breaker EQ PM for Bus 28, 29, 38, 29 Sure Trip Trip

Units

08/26/2025

AR 4895447

2B Medium Range MCR Level Indicator Failing Upscale

09/05/2025

AR 4896013

3-2301-5 Breaker Tripped While Opening

09/08/2025

AR 4898414

U2 CCSW Div 1 Thru Wall Leak

09/17/2025

2E-2351B Sheet

Schematic Diagram Diesel Generator 2/3 Auxiliaries & Start

Relays

AV

2E-2351B sheet

Schematic Diagram Diesel Generator 2/3 Auxiliaries and

Start Relays

AJ

2E-3575BB

Schematic Diagram Control Room Annunciator Panel 903-7

Part 1 of 5

J

2E-3577

Schematic Diagram Main Control Room Annunciator

Chassis Interconnection

E

2E-3815AN

Wiring Diagram Annunciator Input Relay Panel 903-34

Chassis 1-10, TB1-TB10

P

Drawings

2E-6633

Loop Schematic Diagram Reactor Level Instruments

E

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

M-26, Sheet 1

Diagram of Nuclear Boiler and Reactor Recirculating Piping

BT

D2-CCSW-FTG

CCSW System Fatigue Assessment

DRE07-0068

Control Room HVAC Cooling Load Calculation

004

DTP 75

Heat Exchanger Inspection Program

EC 342944

Containment Cooling Service Water (CCSW) System Water

Hammer Loss of Keep-Fill Analysis

EC 349634

Technical Evaluation to Define Control Room Temperature

Parameters and Operation Guidance for Starting Standby

Control Room HVAC System

EC 642308

Use of Unit 3 Boral Coupons for HDFR Corrosion

Surveillance Program

EC 645315

Evaluate for EQ Margin and Increased Qualified Life of

RMS-85-NTC Trip Units Used in 480V Switchgear

08/26/2025

Engineering

Changes

ECR 468871

Out of Tolerance Evaluation

09/15/2025

EQ-66D

EQ Binder: General Electric Type AKD-5 Switchgear

10/24/2017

M12-0-92-002

2/3 DG Power Interlock for Aux.

03/06/1992

Nuclear Event

Report (NER)

NC-10-008-Y

Yellow

Peach Bottom NCV Related to Cable Condition Monitoring

Program

Miscellaneous

QR-033006-8

Qualification Report for Sure-Trip RMS-85 NTC Trip Units

Operability

Evaluations

OP-AA-108-115

Operability Determinations

DAN 902(3)-7 D-

HPCI Turb Supv Trouble

DAN 903-7 C-1

U3 Diesel Gen Clg Wtr PP Trip/Lkout

DIS 1500-14

LPCI System Discharge Header FLow Channel Calibration

and Channel Functional Test

DOP 0400-01

Station Motor Operated Valve Operations

ER-AA-300-150

Cable Condition Monitoring Program

Procedures

MA-AA-716-040

Control of Portable Measurement and Test Equipment

Program

WO 05089230

D2/3 2Y Com CR A/C HX Clean, Insp, Eddy Current Test

09/27/2022

Work Orders

WO 5698073-01

OP Torque Closed HPCI MOV 3-2301-14

07/23/2025

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

AR 4895089

U3 IRM 11 Failed Range 6/7 Overlap Check

09/03/2025

AR 4895179

SRM 21 Reacing HI

09/04/2025

AR 4895883

Insulation in Torus Downcomer

09/07/2025

AR 4896013

3-2301-5 Loss of Light Indication

09/08/2025

Corrective Action

Documents

AR 4896081

Unexpected Alarms: 903-5 D-10, Channel A Rx Scram

09/08/2025

Corrective Action

Documents

Resulting from

Inspection

AR 4895930

NRC ID: Tape on U3 Recirc Motors

09/08/2025

Miscellaneous

D3M24 SSMP

D3M24 Maintenance Outage Shutdown Safety Review

Rev. 1

71111.20

Procedures

DGP 01-01

Unit Startup

211

AR 4556485

ISOL CDSR Bldg Battery Volts (Panel 2223-126D)

2/22/2023

AR 4895380

CRD E-3 Failure to Latch

09/04/2025

AR 4895657

Leaking Air Inlet on 1A MSIV Air Manifold

09/05/2025

AR 4895851

1A MSIV Air Manifold Leak

09/07/2025

Corrective Action

Documents

AR 4896171

Unexpected Alarm 903-3 C-4, ISOL CONDR Temp HI.

09/08/2025

AR 4893091

B ICMU Pump Batt Voltage Not Documented due to

Equipment Issue

08/25/2025

Corrective Action

Documents

Resulting from

Inspection

AR 4898640

NRC ID: A ISO CDSR Make Up PP Room Sump Not

Pumping

09/17/2025

M-25

Diagram of Pressure Suppression Piping

EA

Drawings

M-25 Sheet 1

Diagram of Hardened Containment Ventilation System

B

DMP 0202-01

Recirculation Pump Gen 4 Seal Replacement and Pump

Leak Test

DOP 0040-01

Station Motor Operated Valve Operations

DOP 1300-01

Standby Operation of the Isolation Condenser System

DOS 1400-09

Core Spray System IST Comprehensive/Preservice Pump

Test with Torus Available

DOS 1500-05

2/3A(B) Isolation Condenser Makeup Pump Capacity Test

DOS 4700-02

Drywell Pneumatic Air Flow Qualitative Air Leakage

Assessment

Procedures

MA-AA-716-100

Maintenance Alterations Process

71111.24

Work Orders

WO 01638373

EM D3 10Y PM Replace Control Relay 3-1530-284

05/21/2025

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

WO 05402332

D2 2Y PM Test Diesel Gen #2 Protective Relays/Meters

08/15/2025

WO 05690914

2/3A (B) Isolation Condenser Makeup Pump Quarterly

Operability (DOS 1300-03)

08/21/2025

WO 05719447

EM Replace DCV CRD HCU 18-11 E-3

09/07/2025

WO 1836115

D3 10Y COM MOV Diagnostic Testing & Limitorque

Surv 3-1501-38B

05/27/2025

WO 5382348-01

OPS D2 2Y TS CCSW PMP Comprehensive Oper Test and

IST Surv

07/03/2025

WO 5413098-01

D2 24M TS Torus Area Pri Cnmt Purge/Vent Vlv PS Cal

08/29/2025

WO 5604039-16

OP-PMT-Check 3A Recirc PP Seal for Leaks

09/09/2025

WO 5616288-04

OP-PMT-Check 3B Recirc PP Seal for Leaks at Pressure

09/09/2025

WO 5671225-01

MM-Repair 2-1601-60-SV

08/27/2025

WO 5671225-02

OPS PMT 2-1601-60-SV, U2 Inst Air Supply SV to

AO 2-1601-60

08/28/2025

AR 4895306

3Q25 EP Drill Critique - SCR

09/04/2025

AR 4895312

3Q25 EP Drill Critique - TSC

09/04/2025

AR 4895315

3Q25 EP Drill Critique - OSC

09/04/2025

Corrective Action

Documents

AR 4895319

2Q25 EP Drill Critique - Facilities

09/04/2025

71114.06

Miscellaneous

Dresden 3Q2025 Integrated Drill (TSC Handoff)

08/07/2025

AR 4780725

New Milking Location Found for REMP

06/15/2024

Corrective Action

Documents

AR 4862304

24 ARERR Late

05/01/2025

AR 4867821

NRC ID: Milk Sample Not in the ODCM & Monthly Effluent

Calc

05/22/2025

AR 4867840

NRC ID: REMP Milk Sample Frequency Not in Accordance

with the ODCM

05/22/2025

AR 4867845

NRC ID: Correction to 2024 Ann Effluent Release Rpt

(ARERR)

05/22/2025

Corrective Action

Documents

Resulting from

Inspection

AR 4877104

Correction to 2023 ARERR and AREOR

06/27/2025

24 Annual RGPP Monitoring Report Summary of Results

and Conclusions Dresden Generating Station Morris, Illinois

04/30/2025

Review of the 2024 Land Use Census Data Regarding Goat

Animals Found in the South Sector at 4.2 Miles

07/08/2025

71124.07

Miscellaneous

January

Annual Radiological Environmental Operating Report

05/15/2023

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

through 31

December 2022

January

through 31

December 2023

Annual Radiological Environmental Operating Report

05/15/2024

214338 (1)

Hydrogeologic Investigation Report

11/01/2020

187480-01

D2/3 5Y PM Perform Rad Groundwater Protection

Assessment

10/28/2021

4Q23 QA Report

Teledyne Brown Engineering Environmental Services Quality

Assurance Report - 4th Quarter 2023

01/23/2024

4Q24 QA Report

Teledyne Brown Engineering Environmental Services Quality

Assurance Report - 4th Quarter 2023

01/23/2024

Audit Report

1. 25197

NUPIC Audit of ATI Environmental Inc.

01/10/2023

CY-DR-170-301

Offsite Dose Calculation Manual

EN-AA-408

Radiological Groundwater Protection Program

EN-AA-408-4000

Radiological Groundwater Protection Program

Implementation

Procedures

EN-DR-408-4160

RGPP Reference Material for Dresden Clean Energy Center

04716439-05-02

Self Assessment: Radioactive Solid Waste Processing and

Radioactive Material Handling, Storage, and Transportation

2/05/2024

L100970-4

CFR 61 Waste Stream Analysis: Condensate Resin

09/16/2022

L100970-6

Waste Stream Review and Scaling Factor Determination:

22 ALPS Resin

08/25/2022

L104371-6

CFR 61 Waste Stream Analysis

09/26/2023

NOSA-DRE-24-

Radiation Protection Audit Report

11/01/2024

Miscellaneous

RW-AA-104

Radwaste Storage Facility / Waste Liner Container Integrity

Inspection

04/24/2025

DM-25-001

Kr-85 Source

01/08/2025

DW-24-016

Condensate Liner

09/17/2024

DW-24-024

Alps Resin

07/11/2024

DW-24-044

Torus Filter

2/01/2024

71124.08

Shipping Records

RP-AA-600-1005

Radioactive Waste Shipment Logs

2/11/2024

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

AR 4717507

2-1501-3A

11/14/2023

AR 4770771

2-1501-3A Failed to Open

04/30/2024

AR 4771043

2-1501-3A Valve Lock-Up Caused by Over Range of

Feedback

05/01/2024

AR 4819830

Missed LER for 2-1501-3A

04/30/2024

AR 4838481

MOV 2-1501-3A Crack in Yoke Weld

2/18/2025

Corrective Action

Documents

AR 4841420

IR 4717507 Failure Cause and Screening

2/28/2025

AR 4818810

NRC Walkdown Question

11/20/2024

AR 4902813

NRC ID: WO 5417420 Partial Lost Record

10/01/2025

Corrective Action

Documents

Resulting from

Inspection

AR 4907553

NRC ID: WO 1082615-01 2A LPCI DP Isolator Not in EDMS

09/29/2025

71152A

Work Orders

WO 05417420

2-1501-3A Did Not Open

11/30/2023

AR 4770771

2-1501-3A Failed to Open

04/30/2024

Corrective Action

Documents

AR 4885552

Troubleshooting Actions Not Timely

07/24/2025

WO 05417420

2-1501-3A Did Not Open

11/30/2025

WO 05550335

EMD Adjust MOV 2-1501-3A Limits/Stroke Length

2/05/2024

71152S

Work Orders

WO 99212153

D2 29Y EQ LPCI 1501-3A VLV Potentiometer Replacement

04/04/2022

71153

Corrective Action

Documents

Resulting from

Inspection

AR 4897297

NRC ID: Update Needed to LER 249/2024-002-01

09/08/2025