IR 05000315/2025001
| ML25125A150 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/07/2025 |
| From: | Nestor Feliz-Adorno NRC/RGN-III/DORS/ERPB |
| To: | Ferneau K Indiana Michigan Power Co |
| References | |
| IR 2025001 | |
| Download: ML25125A150 (1) | |
Text
SUBJECT:
DONALD C. COOK NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000315/2025001 AND 05000316/2025001
Dear Kelly Ferneau:
On March 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Donald C. Cook Nuclear Plant. On April 22, 2025, the NRC inspectors discussed the results of this inspection with S. Dailey, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at Donald C. Cook Nuclear Plant.
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 Donald C. Cook Nuclear Plant.
May 7, 2025 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, Néstor J. Féliz Adorno, Chief Engineering and Reactor Projects Branch Division of Operating Reactor Safety Docket Nos. 05000315 and 05000316 License Nos. DPR-58 and DPR-74 Enclosure:
As stated cc: Distribution via LISTSERV Signed by Feliz-Adorno, Nestor on 05/07/25
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Donald C. Cook Nuclear Plant, 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 Correct Degraded Power Range Nuclear Instrumentation Potentiometers Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000315,05000316/2025001-01 Open/Closed
[P.2] -
Evaluation 71111.12 A self-revealed finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B,
Criterion XVI, Corrective Actions was identified for the licensees failure to assure that conditions adverse to quality are promptly corrected. Specifically, the licensee failed to correct degraded ion current gain potentiometers for Unit 2 Power Range Nuclear Instrumentation Channel IV, 2-NRI-44. This resulted in multiple alarms in the control room, a partial trip signal, and automatic control rod movement.
Inadequate Post-Maintenance Test Led to an Overpower Event Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000315/2025001-02 Open/Closed
[H.14] -
Conservative Bias 71153 A self-revealed finding of very low safety significance (Green) and associated non-cited violation (NCV) of the Donald C. Cook, Unit 1 Renewed Facility Operating License,
Condition 2.C.(1), Maximum Power Level, was identified for the failure to operate with reactor power not exceeding 3304 megawatts thermal. Specifically, an inadequate post-maintenance test for the Unit 1 Turbine Control Valve 2 (CV-2) linear voltage differential transformer (LVDT) resulted in an incorrectly manufactured LVDT being installed. The LVDT positioned CV-2 full open, causing a power transient.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000316/2024-004-00 LER 2024-004-00 for Donald C Cook Nuclear Plant, Unit 2, 2AB Emergency Diesel Generator Inoperable for Longer Than Allowed by Technical Specifications 71153 Closed
LER 05000316/2024-004-01 LER 2024-004-01 for Donald C. Cook Nuclear Plant, Unit 2, 2AB Emergency Diesel Generator Inoperable for Longer Than Allowed by Technical Specifications 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On March 19, 2025, Unit 1 reduced power to approximately 63 percent and again on March 21, 2025, to approximately 45 percent of rated thermal power for main steam safety valve testing. The unit was shut down on March 22, 2025, for a planned refueling outage and remained offline for the rest of the inspection period.
Unit 2 began the inspection period at rated thermal power. On March 7, 2025, Unit 2 reduced power to approximately 90 percent of rated thermal power for main turbine valve testing. The unit returned to full rated thermal power early the next morning and remained at or near full rated thermal power for the rest of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html.
Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk-significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather associated with multiple days of below zero temperatures on January 21, 2025.
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)1AB emergency diesel generator (EDG) during 1CD EDG work window on January 28, 2025
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 Zones 15/16/18/19/79/80/84/85: AB/CD diesel generator room and turbine room (NE and SE portion), both units, elevation 587'-0" and 591'-0" on January 16, 2025
- (2) Fire Zones 44C/D/G/H: E/W residual heat exchanger room, both units, elevation 609'-0" on February 7, 2025
- (3) Fire Zones 40A/40B/41: 4kV AB/CD switchgear rooms and engineering safety system rooms, Unit 1, elevation 609'-6" on March 27, 2025
- (4) Fire Zones 42A/B/C/D: emergency power systems areas, Unit 1, elevation 609'-6" on March 27, 2025
- (5) Fire Zone 144: hot shutdown panel enclosure for Unit 1 on March 27, 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 February 6, 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) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during reactor shutdown and plant cooldown for Unit 1 refueling outage on March 22, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated just-in-time simulator training associated with refueling outage tasks on March 19, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Unit 2 N-44 Power Range Channel and its 10 CFR 50.65 a
- (1) consideration
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) Risk management during 2-QRV-251, Unit 2 chemical volume and control system (CVCS) charging pumps discharge flow control valve emergent work window on January 22, 2025
- (2) Configuration Risk Management Program during 1CD EDG maintenance work window on January 28, 2025
- (3) Risk management during 1E essential service water (ESW) pump maintenance work window on February 3, 2025
- (4) Risk management during Unit 1 lowered inventory operations for refueling outage U1C33 on March 28, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (3 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
- (2) AR 2025-1405; 1-ERS-7401; Unit 1 Control Room Radiation Monitor, reading software fault on February 21, 2025
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Replacement of chemical volume and CVCS cross tie isolation valves (1-CS-534, 1-CS-535, and 1-CS-536)
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Partial)
(1)
(Partial) The inspectors evaluated refueling outage U1C33 activities from March 22, 2025, to March 31, 2025.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (4 Samples)
- (1) Unit 1 east motor-driven auxiliary feedwater pump discharge valves, 1-FMO-222 and 1-FMO-232, stroke test following external preventative maintenance on January 6, 2025 (2)2-QRV-251, CVCS charging pumps discharge flow control valve, post-maintenance testing following valve repack on February 12, 2025 (3)1CD EDG post-maintenance testing following planned work window on February 14, 2025
- (4) East diesel-driven fire pump post-maintenance testing following leak repairs on March 17, 2025
Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) Unit 2 west centrifugal charging pump surveillance run on January 16, 2025
- (2) 1E component cooling water surveillance on March 5, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) Unit 2 west containment spray system operability and Group B pump test on January 31,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===
- (1) Unit 1 (January 1 through December 31, 2024)
- (2) Unit 2 (January 1 through December 31, 2024)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
- (1) Unit 1 (January 1 through December 31, 2024)
- (2) Unit 2 (January 1 through December 31, 2024)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
- (1) Unit 1 (January 1 through December 31, 2024)
- (2) Unit 2 (January 1 through December 31, 2024)
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04) (2 Samples)
- (1) Unit 1 (January 1 through December 31, 2024)
- (2) Unit 2 (January 1 through December 31, 2024)
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)
- (1) The inspectors evaluated the unexpected opening of a turbine control valve and licensees performance on November 21, 2024.
Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000316/2024-004-00, 2AB Emergency Diesel Generator Inoperable for longer than allowed by Technical Specifications (ADAMS Accession No.ML24316A006) and LER 05000316/2024-004-01, 2AB Emergency Diesel Generator Inoperable for longer than allowed by Technical Specifications (ADAMS Accession No. ML25086A101).
The inspection conclusions associated with these LERs are documented in Special Inspection Reactive Report 05000315/2024050 and 05000316/2024050 under the Inspection Results Section for the finding and violation titled, Failure to Identify and Correct the Cause of May 2024 Failed 2 AB EDG Slow Speed Start Surveillance, (ADAMS Accession No. ML25027A426). These LERs are Closed.
Personnel Performance (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the unexpected opening of a power-operated relief valve on a main steam line and licensees response on March 11,
INSPECTION RESULTS
Failure to Correct Degraded Power Range Nuclear Instrumentation Potentiometers Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000315,05000316/2025001-01 Open/Closed
[P.2] -
Evaluation 71111.12 A self-revealed finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Actions was identified for the licensees failure to assure that conditions adverse to quality are promptly corrected. Specifically, the licensee failed to correct degraded ion current gain potentiometers for Unit 2 Power Range Nuclear Instrumentation Channel IV, 2-NRI-44. This resulted in multiple alarms in the control room, a partial trip signal, and automatic control rod movement.
Description:
On June 20, 2024, safety-related Unit 2 Power Range Nuclear Instrumentation Channel IV, 2-NRI-44, spiked high. The licensees investigation determined the direct cause to be a momentary loss of continuity in the gain potentiometers in the channel drawer due to the buildup of hardened grease. The licensee consulted with the vendor, who verified that the multiple multi-turn potentiometers within the drawer are all susceptible to the degraded condition. Consequently, AR 2024-5163 identified that all the potentiometers in the instrument channel drawer were susceptible. However, the corrective action that was created specified to wipe only the channel gain potentiometers. There was no discussion about how to address the condition in the other potentiometers, nor why they were not included. A preventative maintenance task was performed on July 3, 2024, under Work Order (WO) C10073010007 to wipe potentiometers for 2 NRI 44 by cycling them up and down several times to clean the contact area. The task stated Adjust/wipe potentiometers on 2-NRI-44A-DWR &
2-NRI-44B-DWR as necessary per Operations direction. When the task was performed, only the channel fine gain potentiometer on the front of the channel drawer was wiped at that time.
On August 11, 2024, 2-NRI-44 spiked high again. This time it produced multiple alarms in the control room, generated a partial trip signal, and resulted in automatic control bank D rod movement to reduce reactor power. The invalid signal indicated that power had exceeded 100 percent, causing control rods to insert. Rods were inserted from 224.5 to 219 steps before operator action placed rod control in manual. Operators placed 2-NRI-44 in trip and returned control bank D rods to the original step demand at 224.5 steps. Placing the channel in trip changed the protection logic from a 2-out-of-4 logic to a 1-out-of-3, increasing the vulnerability to a reactor trip from any additional trip signal, whether valid or invalid.
Upon completion of the investigation of the failure of 2-NRI-44 on August 11, 2024, the licensee determined that the direct cause was an erratic signal caused by the failure of the ion current gain potentiometers. The licensee concluded that the corrective actions taken in response to the June 20, 2024, event were not sufficient, as the ion current gain potentiometers should also have been wiped. Additionally, the site categorized this event as a Maintenance Preventable Functional Failure due to the repeated occurrence of 2-NRI-44 spiking high as indication of a failure to correct the cause of the event.
Corrective Actions: The licensee took immediate corrective actions by replacing the failed potentiometers on August 11, 2024. Planned corrective actions going forward include clarifying which potentiometers need to be wiped and how often. Additionally, the licensee will consider evaluating the need for preventive maintenance activities to replace potentiometers on a frequency basis.
Corrective Action References: AR 2024-5163, Unit 2 N44 Power Range Spike, and AR 2024-6194, U2 N-44 Failed High
Performance Assessment:
Performance Deficiency: The licensees failure to correct degraded safety-related ion current gain potentiometers for 2-NRI-44, a condition adverse to quality, 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 Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to properly correct the degraded ion current gain potentiometers for 2-NRI-44 resulted in an invalid high signal, which caused unintended control rod insertion of 5.5 steps, altering core reactivity and plant stability.
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 determined that this finding is of very low safety significance (Green) using Exhibit 1, Initiating Events, Section B, Transient Initiators, because it neither caused a reactor trip nor resulted in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, although AR 2024-5163 identified that all the potentiometers associated with 2-NRI-44 were susceptible to the degraded condition that was determined to be the most likely cause of the spiking, the licensee did not evaluate their choice of wiping only the channel gain potentiometers. This resulted in the ion current gain potentiometers not being corrected.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures be established to assure that conditions adverse to quality, such as failures, malfunctions, and deficiencies are promptly corrected.
Contrary to the above, between June 20 and August 11, 2024, the licensee failed to promptly correct a condition adverse to quality. Specifically, after the licensee identified that all the potentiometers in the safety-related 2-NRI-44 channel drawer were susceptible to a buildup of hardened grease, the licensees work order to wipe the potentiometers during maintenance on July 3, 2024, only addressed the degraded condition on the fine gain potentiometer. The licensee closed the work order with no future actions planned for the remaining potentiometers. Consequently, the degraded condition on the ion current gain potentiometers resulted in a power transient on August 11, 2024.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Inadequate Post-Maintenance Test Led to an Overpower Event Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000315/2025001-02 Open/Closed
[H.14] -
Conservative Bias 71153 A self-revealed finding of very low safety significance (Green) and associated non-cited violation (NCV) of the Donald C. Cook, Unit 1 Renewed Facility Operating License, Condition 2.C.(1), Maximum Power Level, was identified for the failure to operate with reactor power not exceeding 3304 megawatts thermal. Specifically, an inadequate post-maintenance test for the Unit 1 Turbine Control Valve 2 (CV-2) linear voltage differential transformer (LVDT) resulted in an incorrectly manufactured LVDT being installed. The LVDT positioned CV-2 full open, causing a power transient.
Description:
On November 6, 2024, operators in the Unit 1 control room noted that the Servo Position Controller (SPC) for Turbine Control Valve 4 (CV-4) had automatically swapped from Train A to its redundant Train B following a turbine load change. A system engineer was contacted to investigate the cause and found a fault on the coils controlled by the Train A SPC.
The licensee conducted an extent of condition to determine if any other redundant SPCs on the remaining three control valves had failed. Operators were tasked with manually swapping between the SPCs on each control valve to observe if the system automatically swapped back to the previous SPC. In doing so, the redundant SPC would attempt to energize its coil and indicate a fault if unable to do so. Operations assumed any fault would cause the software to automatically select back to the known functional train.
On November 21, 2024, while performing this extent of condition testing at full power, operators in the Unit 1 control room selected the standby Train B SPC for CV-2. During this test, the valve began opening automatically, causing an increase in reactor power. Reactor power exceeded the licensed maximum thermal power limit defined in the Donald C. Cook Unit 1 Renewed Facility Operating License, Condition 2.C.(1), Maximum Power Level, of 3304 megawatts thermal. Instantaneous reactor peaked at 102.78 percent as read by the plant computer. The reactor remained above 100 percent licensed power for approximately 7 minutes while operators followed their procedures to reduce power in a controlled manner.
Following the event, it was discovered that the LVDTs for both trains on CV-2 had been recently replaced on November 13, 2024, under Work Order (WO) C10078403, for unrelated reasons. The post-maintenance test (PMT), per WO C10078403 Step 5, Post Maintenance Testing, directed maintenance personnel to verify that LVDT B reflects current plant conditions. Maintenance staff performed a channel check with the redundant Train A to ensure the two agreed. This revealed a 1.5-inch discrepancy in mounting location. A cross-functional team consisting of engineering, maintenance, and operations concluded that the difference in indication was a calibration issue and planned to correct it during the next outage, when full control valve testing would occur.
These actions were contrary to licensee procedure PMI-2294, Post Maintenance Testing Program. Paragraph 4.2.2(a) requires, in part, that planning shall consult with plant engineering as necessary to determine the correct PMT. The procedure states that PMTs are intended to confirm that structures, systems, and/or components (SSCs) are capable of performing their intended functions when returned to service following maintenance/repair, and to ensure that no new or related deficiencies have been created by the maintenance/repair activity.
Subsequent bench testing revealed that the Train B LVDT was wired backwards. The licensee had procured this non-safety-related LVDT from another licensee. It was determined that the vendor supplying these LVDTs had switched to a new manufacturer whose internal coil winding practices differed due to use in other industries. These LVDTs were inadvertently supplied to the nuclear industry.
On December 9, 2024, the NRC completed its evaluation of the November 21, 2024, event to determine whether a reactive inspection was warranted. The NRC concluded that a reactive inspection was not warranted. The evaluation can be accessed in ADAMS under Accession No. ML24344A106.
Corrective Actions: The licensee replaced the Train B LVDT with one that had previously undergone control valve testing, which verified its full range of indication. They placed a hold on all LVDTs in stock and plan to functionally test them. In response to the event, they performed a condition evaluation on fuel margins and conducted an equipment apparent cause evaluation of LVDT PMTs. Additionally, they revised the site PMT guidance procedure to specify actions for online LVDT replacement, including observing the full usable range before placing the component in service.
Corrective Action References: ARs 2024-8641, 2024-8643, 2025-1283
Performance Assessment:
Performance Deficiency: The licensees failure to determine the correct PMT for the Unit 1 Turbine Control Valve 2 (CV-2) LVDT was contrary to Revision 9 of procedure PMI-2294, Post Maintenance Testing Program, Paragraph 4.2.2(a), and was a performance deficiency.
Specifically, CV-2 had its Train B LVDT replaced online with an incorrectly manufactured one and the PMT did not confirm the LVDT was capable of performing its intended function when returned to service following the maintenance activity.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to determine the correct post-maintenance test following replacement of the Train B LVDT for CV-2 resulted in an unanticipated secondary side transient when swapping to that train and caused a power transient that exceeded the licensees licensed maximum power level.
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 determined this finding was of very low safety significance (Green), because they answered No to the Initiating Events screening question in Exhibit 1, Section B, Transient Initiators.
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, while working the PMT for the linear voltage differential transformer, it was identified that the as-left conditions were different from the other installed one. The adequacy of the PMT was rationalized even when information was incomplete, and conditions were unusual.
Enforcement:
Violation: Donald C. Cook Unit 1 Renewed Facility Operating License, Condition 2.C.(1),
Maximum Power Level, states that the licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3304 megawatts thermal.
Contrary to the above, on November 21, 2024, from 10:26 a.m. to 10:33 a.m. EST, the licensee failed to operate the facility (Unit 1) at steady state reactor core power levels not exceeding 3304 megawatts thermal. Specifically, due to an incorrectly wired LVTD in a control valve, reactor thermal power exceeded the licensed limit of 3304 megawatts thermal (100 percent licensed power), peaking at 102.78 percent as indicated by the plant computer.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 22, 2025, the inspectors presented the integrated inspection results to S. Dailey, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
AR 2025-0083
Cold Spot Behind I-Beam Near Unit 1 East FRV Gallery
01/05/2025
AR 2025-0234
Freezing Air Intrusion into the Auxiliary Building
01/08/2025
AR 2025-0463
Screenhouse Roll Up Door Not Working Correctly
01/16/2025
Corrective Action
Documents
AR 2025-0519
U1 Fire in the W FRV Gallery on a Heater Cable Connector
01/19/2025
2-IHP-5040-
EMP-004
Plant Winterization and De-winterization
2-OHP-4022-
001-010
Severe Weather
Procedures
PMP-5055-SWM-
001
Severe Weather Guidelines
OP-1-5151A-51
Flow Diagram Emergency Diesel Generator AB Unit No. 1
Drawings
OP-1-5151B-65
Flow Diagram Emergency Diesel Generator AB Unit #1
Procedures
1-OHP0-4021-
2-008AB
Operating DG1AB Subsystems
AR 2010-7176
Fire Door Will Not Close Due To Air Pressure
01/21/2011
AR 2025-0195
Missed Required Actions of TRM
01/07/2025
AR 2025-0309
1-DQR-CO2-RBX Light Bulb Base Failed
01/09/2025
Corrective Action
Documents
AR 2025-2016
PMP-2270-CCM-001 Violation / Missed Combustible Watch
03/17/2025
FIRE-PRE-
PLANS-
VOLUME-1
Fire Pre-Plans Volume 1
Fire Plans
D.C. Cook Nuclear Plant Fire Safety Analysis (FSA)
25-060-A
Fire Drill No: 125-060-A PZR Transformer Room
2/06/2025
TCP-2025-069
Transient Combustible Materials Permit - 4kV Transformer
Room 609' Turbine
03/11/2025
Miscellaneous
TCP-2025-082
Transient Combustible Materials Permit - 4kV Room at
Penetration W3073
03/17/2025
Procedures
2-FPP-4030-
066-026
Technical Requirements Manual Fire Door Inspection
Miscellaneous
RQ-J-0205
U1C33 Shutdown JITT Plan
Procedures
1-OHP-4021-001-
003
Power Reduction
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
1-OHP-4021-001-
004
Plant Cooldown from Hot Standby to Cold Shutdown
24-5163
Unit 2 N44 Power Range Profile
06/20/2024
Corrective Action
Documents
24-6194
U2 N-44 Failed High
08/11/2024
Miscellaneous
VTD-WEST-1004
Westinghouse NSD Technical Bulletin for Multi-Turn
Potentiometers [Pub. # NSD-TB-20]
04/20/2020
Procedures
PMP-2291-MM-
001
Minor/Tool Pouch Maintenance
Work Orders
C1007301007
MTI, 2-NRI-44A/B-DWR, Adjust/Wipe Potentionmeters as
Needed
11/29/2024
AR 2013-12888
Clarification on TS 3.8.1B
09/01/2013
AR 2024-7669
2-QRV-251 Packing Leak
10/11/2024
Corrective Action
Documents
AR 2025-0575
2-QRV-251 Packing Leak
01/20/2025
Corrective Action
Documents
Resulting from
Inspection
AR 2025-0662
2-CS-628 is Leaking Past its Pipe Cap
01/23/2025
Drawings
OP-2-5129-59
Flow Diagram CVCS-Reactor Letdown and Charging
Unit No. 2
U1C33 Refueling Outage Shutdown Safety Plan Report
03/06/2025
Plan of the Day Meeting
2/03/2025
Plan of the Day Meeting Package
03/27/2025
Miscellaneous
Plan of the Day Meeting
01/28/2025
1-OHP-4030-114-
21
Event Initiated Surveillances
1-OHP-4030-114-
031
Operations Weekly Surveillance Checks
2-OHP-4030-
033-001
Supplemental Diesel Generator Testing
2-OHP-4021-003-
001
Letdown, Charging and Seal Water Operation
PMP-2291-OLR-
001
On-Line Risk Management
Procedures
PMP-2291-OLR-
On-Line Risk Management
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
001
PMP-4030-EIS-
001
Event Initiated Surveillance Testing
PMP-4100-SDR-
2
Outage Risk Assessment and Management
Calculations
MD-12-ESW-107-
N
ESW Make-up Flow to EDG Jacket Water System
AR 2024-9257
2/19/2024
Corrective Action
Documents
AR 2025-1405
1-ERS-7401 Reading SOFTWARE FAULT
2/21/2025
AEP:NRC:0398R
Application for Amends to Licenses DPR-58 & DPR-74,
modifying Control Room Emergency Ventilation Sys &
Associated Bases
06/29/1989
AEP:NRC:0914E
Modification of Submittals Regarding Control Room
Ventilation
2/29/1988
Miscellaneous
Amendment No.
159
Docket No. 50-315, Donald
- C. Cook Nuclear Plant, Unit
No. 1, Amendment to Facility Operating License
11/20/1991
Operability
Evaluations
AR 2024-9165-1
Operability Determination Supplement for AR 2024-9165;
Additional Analysis Needed For Already Identified Condition
2/19/2024
Procedures
1-OHP-4024-119
Annunciator #119 Response: Station Auxiliary AB
Engineering
Changes
Upgrade CVCS Crosstie Valves 1-CS-534, 1-CS-535, & 1-
CS-536 To Reduce Crosstie Leakage
Corrective Action
Documents
AR 2025-2585
2-FTPL E08 Conveyor Drive Faults Experienced during
Offload
03/31/2025
Miscellaneous
U1C33 Refueling Outage Shutdown Safety Plan Report
03/06/2025
1-OHP-4022-002-
006
Loss of Refueling Water Level during Refueling Operations
1-OHP-4022-016-
004
Loss of Component Cooling Water
1-OHP-4022-017-
001
Loss of RHR Cooling
1-OHP-4030-001-
2
Containment Inspection Tours
Procedures
1-OHP-4030-114-
010
Containment Isolation
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
PMI-4100
Shutdown Risk Management
PMP-4100-SDR-
001
Plant Shutdown Safety and Risk Management
PMP-4100-SDR-
2
Outage Risk Assessment and Management
AR 2021-7256
1-CCW-120 Failed to Fully Stroke During Surveillance
2/05/2023
AR 2024-6455
Repack 2-QRV-251
08/22/2024
AR 2024-7465
1CD EDG Engine Analysis Indications
10/08/2024
AR 2024-7669
2-QRV-251 Packing Leak
10/11/2024
AR 2024-8485
Loose Bolt on 1CD EDG Exhaust Support Bracket
11/14/2024
AR 2025-0575
2-QRV-251 Packing Leak
01/20/2025
AR 2025-1602
2-PP-145E Pump Casing Leak
03/01/2025
Corrective Action
Documents
AR 2025-1728
Battery Connection Resistance Checks Not Performed
03/06/2025
Corrective Action
Documents
Resulting from
Inspection
AR 2025-0773
NRC Review of Completed MOV External PMs
01/28/2025
OP-1-5106A-68
Flow Diagram Aux-Feedwater Unit 1
OP-1-5135-44
Flow Diagram CCW Pumps and CCW Heat Exchangers
05/02/2019
OP-1-5135A-47
Flow Diagram CCW Safety Related Loads
07/15/2021
Drawings
OP-2-5129-72
Flow Diagram CVCS-Reactor Letdown & Charging Unit No 2
1-OHP-4021-032-
001CD
DG1CD Operation
1-OHP-4030-116-
20E
East Component Cooling Water Loop Surveillance Test
1-OHP-4030-132-
27CD
1: DG1CD Slow Speed Start
1-OHP-4030-156-
017E
Manual Full Stroke Exercise Test Of 1-FMO-222 and
1-FMO-232
1-OHP-4030-156-
017E
East MDAFP Valve Lineup
Procedures
2-IHP-5021-
EMP-008
Battery Connection Maintenance
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2-OHP-4030-
066-121FD
Diesel Fire Pump Operability Test
2-OHP-4021-003-
001
16: Shifting Charging Flow Control to and from
Alternate Flow Path
2-OHP-4030-203-
2W
West Centrifugal Charging Pump Operability Test
2-OHP-4030-209-
007W
West Containment Spray System Test
EHI-5071
Inservice Testing Program Implementation
OHI-4016
Data Sheet 3: IST Valve Operability After Maintenance
PMI-5071
Inservice Testing
TDB-1-FIG-19-8
Safety Related Throttled Valves
C10065562022
1-OME-150-CD, Component/Electrical/Air Gap Check
01/30/2025
C10067119
East CCW Full-Stroke Exercise Tests: Manually Operated
Valves
03/12/2025
C10073631001
1-FMO-232, PERFORM EXTERNAL PM
01/06/2025
C10073632001
1-FMO-222 PERFORM EXTERNAL PREVENTIVE
MAINTENANCE
01/06/2025
C10074782006
2-QP-92E, Replace Leaking Pipe Nipple
03/05/2025
C10077028001
2-QRV-251; Repack Valve with 5718/5000 Packing
01/22/2025
C10077028002
2-QRV-251; PMT Leak Inspection
01/23/2025
Work Orders
C10077805
West Centrifugal Charging Pump Group A Test
01/20/2025
Corrective Action
Documents
AR 2025-1317
Missed Safety System Functional Failure Report on U2
2/18/2025
Operational Narrative Logs
01/01/2024 -
2/31/2024
Miscellaneous
AEP-NRC-2025-
D.C. Cook Units 1 and 2 - 4Q2024 - PI Data Elements
Change Report (CR)
03/19/2025
71151
Procedures
PMP-7110-PIP-
001
Reactor Oversight Program Performance Indicators and
Monthly Operating Report Datasheets (January 2024 thru
December 2024)
24-5817-5
EACE for 2AB EDG Failed Surveillance (Freq Low).
01/22/2025
24-6894
Potential Past Operability Concern on 2AB
03/19/2025
Corrective Action
Documents
24-7274
Ineffective Evaluation
03/06/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
24-8300
U1 MT CV-4 SPC Swapped from SPC-A to SPC-B after
Load Change
01/23/2025
24-8641
SPC Failure while Swapping on CV2 in Unit 1
01/23/2025
24-8643
Hourly LEFM RX Power Over 100% during Transient
01/23/2025
25-1878
2-RU-15 SG4 PORV 2-MRV-243 CONTROLLER
03/19/2025
Unit 2 Control Room Narrative Log
03/11/2025
Miscellaneous
ECP 1-T2-07
Unit 1 Main Turbine TS3000 Digital Control System Basis of
Design Document
1-OHP-5030-050-
001
Main Turbine and Feed Pump Turbine Valve Functional
Checks
2-OHP-4022-IFR-
001
Instrument Failure Response
OHI-4000
Conduct of Operations: Standards
156
PMI-2294
Post-Maintenance Testing Program
Procedures
PMP-2291-PMT-
001
Work management Post Maintenance Testing Matrices
Work Orders
C10078403
(2024-8226) U1 MT CV 2 Positioner B Box Vibration
11/14/2024