IR 05000237/2025010
| ML25119A186 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/30/2025 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| IR 2025010 | |
| Download: ML25119A186 (1) | |
Text
SUBJECT:
DRESDEN NUCLEAR POWER STATION - BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000237/2025010 AND 05000249/2025010
Dear David Rhoades:
On March 21, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at your Dresden Nuclear Power Station and discussed the results of this inspection with H. Patel, Plant Manager and other members of your staff. The results of this inspection are documented in the enclosed report.
The NRC inspection team reviewed the stations problem identification and resolution program to confirm that the station was complying with NRC regulations and licensee standards. Based on the samples reviewed, the team determined that your program complies with NRC regulations and applicable industry standards such that the Reactor Oversight process can continue to be implemented.
The team also evaluated the stations effectiveness in identifying, prioritizing, evaluating, and correcting problems, reviewed licensee audits and self-assessments, and its use of industry and NRC operating experience information. The results of these evaluations are in the enclosure.
Finally, the team reviewed the stations programs to establish and maintain a safety-conscious work environment and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews, the team found no evidence of challenges to your organizations safety-conscious work environment. Your employees appeared willing to raise nuclear safety concerns through at least one of the several means available.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional April 30, 2025 Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at Dresden Nuclear Power Station.
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.
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
Inspection Report
Docket Numbers:
05000237 and 05000249
License Numbers:
Report Numbers:
05000237/2025010 and 05000249/2025010
Enterprise Identifier:
I-2025-010-0052
Licensee:
Constellation Energy Generation, LLC
Facility:
Dresden Nuclear Power Station
Location:
Morris, IL
Inspection Dates:
March 03, 2025 to March 21, 2025
Inspectors:
K. Henry, Regional Governmental Liaison Officer
M. Porfino, IEMA Resident Inspector-Dresden
N. Shah, Senior Project Engineer
C. St. Peters, Senior Project Engineer
Approved By:
Robert. Ruiz, Chief
Reactor Projects Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a biennial problem identification and resolution inspection at Dresden Nuclear Power Station, 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 Identify a Condition Adverse to Quality with the Local Power Range Monitors Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237,05000249/2025010-01 Open/Closed
[P.1] -
Identification 71152B The inspectors identified a Green finding and associated Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, when the licensee failed to identify as a Condition Adverse to Quality (CAQ), that the transformers used in the local power range monitor (LPRM) power supplies were of unanalyzed design, and therefore, a potential operability issue affecting the safety-related average and oscillation power range monitoring systems.
Additional Tracking Items
None.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
OTHER ACTIVITIES - BASELINE
71152B - Problem Identification and Resolution Biennial Team Inspection (IP Section 03.04)
- (1) The inspectors performed a biennial assessment of the effectiveness of the licensees Problem Identification and Resolution program, use of operating experience, self-assessments and audits, and safety-conscious work environment.
- Problem Identification and Resolution Effectiveness: The inspectors assessed the effectiveness of the licensees Problem Identification and Resolution program in identifying, prioritizing, evaluating, and correcting problems. The inspectors also conducted a five-year review of the electrical distribution system.
- Operating Experience: The inspectors assessed the effectiveness of the licensees processes for use of operating experience.
- Self-Assessments and Audits: The inspectors assessed the effectiveness of the licensees identification and correction of problems identified through audits and self-assessments.
- Safety-Conscious Work Environment: The inspectors assessed the effectiveness of the stations programs to establish and maintain a safety-conscious work environment.
INSPECTION RESULTS
Assessment 71152B Assessment of the Corrective Action Program Overall, the team concluded that the licensee had established a low threshold for entering deficiencies into the corrective action program (CAP), that the issues were generally being appropriately prioritized and evaluated for resolution, and that corrective actions (CAs) were implemented to mitigate the future risk of issues occurring that could affect overall system operability and/or reliability.
Effectiveness of Problem Identification
Overall, the station was effective at identifying issues at a low threshold and was properly entering them into the CAP as required by station procedures. Workers were familiar with how to enter issues into the CAP and stated that they were encouraged to use it to document issues. During plant walkdowns, the team observed that issues were being identified in the field and that they were being properly addressed in the CAP. The team determined that the station was generally effective at identifying negative trends that could potentially impact nuclear safety.
The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV), for the failure to identify a condition adverse to quality (CAQ) via the CAP, associated with the safety-related local power range monitors (LPRM). This issue is discussed in more detail later in the report.
Effectiveness of Prioritization and Evaluation of Issues
The inspectors noted that issues were properly screened with most either classified as CAQs or Non-Corrective Action Program (NCAP) items. Through a selective review of CAP and NCAP items, the inspectors found no issues either with the assigned level of evaluation or the proposed CAs. Issues having potential operability concerns were properly addressed through the screening process and during control room observations and accompaniment of non-licensed operators during daily rounds. The inspectors did not identify any significant operator workarounds or similar deficiencies.
The inspectors did a selective review of issues identified by the NRC either documented as observations or for which findings or other enforcement was issued. These issues were properly documented and screened in the CAP and corrective actions were appropriate and timely scheduled.
Issue evaluations were generally sound and of good quality. Most issues were screened as low significance and were assigned a work group evaluation (the lowest level of review);more significant issues were assigned a Corrective Action Program Evaluation (CAPE) or if highly significant, a root cause evaluation. Through a selective review, the inspectors verified that the assigned evaluations were consistent with the significance of the issue as defined in the licensees process.
Effectiveness of Corrective Actions
The inspectors reviewed issues in the CAP to ensure appropriate classification and prioritization of the problems resolution commensurate with the safety significance of the issue. Corrective actions were assessed to ensure they were appropriately focused to correct the problem identified and to address the root and contributing causes of significant conditions adverse to quality and conditions adverse to quality. The inspectors reviewed completion of CAs to validate they were completed according to the action plan, in a timely manner, and were effective at addressing the issue and preventing future issues. For NRC-identified issues, the inspectors evaluated whether prior attempts by the licensee to remedy the problems were adequate.
Based on the samples reviewed, the team determined that the licensee was generally effective in corrective action implementation. Problems identified using a root cause or other cause methodologies were resolved in accordance with CAP requirements.
One finding with an associated non-cited violation was identified.
Assessment 71152B Assessment of Operating Experience The licensee routinely screened industry and NRC Operating Experience (OPEX) information for station applicability. Based on these initial screenings, the licensee assigned actions in the CAP to evaluate the potential station impact. During interviews, licensee staff stated that operating experience lessons-learned were communicated during work briefings and department meetings and incorporated into plant operations.
The OPEX Review template states that If there are no Operability Concerns, mark this section as N/A with a short answer as to why. The inspectors noted that for OPXR ATI Assignment #4671637-02, section three, operability concerns, N/A was selected but no reason was listed as to why. Subsequently, licensee staff stated that the OPEX was evaluating the applicability of an administrative vulnerability and there was no system or equipment operation evaluated in the review; however, the failure to follow the template was not discussed in the response. The improper completion of the OPEX review was documented as IR 4846746. The inspectors also noted that the operating experience was appropriately reviewed as a potential precursor during CAPE and root cause evaluations.
No findings or violations were identified.
Assessment 71152B Assessment of Licensee Self-Assessments and Audits The inspectors reviewed several audits and self-assessments and deemed those sampled as thorough and intrusive with regards to following up with the issues that were identified. The Maintenance Rule Database was not updated to reflect an April 2023 change to the classification for the U2/U3 low pressure coolant injection/containment cooling service water (LPCI/CCSW) function 15-6 from low to high safety significance. This update was required by Step 4.2.2.2 of ER-AA-320, Maintenance Rule Implementation per NEI 18-10. This was also a missed opportunity, as the most recent assessment of the maintenance rule program (AR 2466136-91 Maintenance Rule Periodic Assessment #15 ) mentioned the change in the function significance, but did not identify the database discrepancy. The failure to follow procedure ER-AA-320, was considered a minor violation, as the Team did not identify any examples where a maintenance rule issue involving this function had been inadequately assessed since the change was made. This was also consistent with example 7.d of NRC Inspection Manual Chapter 612, Appendix E, Examples of Minor Issues, dated October 26, 2023. The licensee documented this issue as IR 4846830, with an assigned action to verify that LPCI/CCSW maintenance rule issues occurring after the 2023 change were properly evaluated.
One minor violation was identified.
Assessment 71152B Assessment of the Safety-Conscious Work Environment The team reviewed the stations programs to establish and maintain a safety-conscious work environment and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews, workers at the station expressed freedom to raise and enter safety concerns through any one of the various avenues available to them, and the team encountered no indications of a chilled work environment.
Workers expressed favorable opinions of the Employee Concerns Program (ECP) during interviews. While most workers felt no need to engage the ECP, the inspectors noted that there were still several issues documented in the program. Through a selective review, the inspectors concluded that these issues were appropriately handled and identified no adverse trends. The inspectors did note, however, that some radiation protection staff were unaware of the ECP, in spite of several signs posted throughout the station describing the ECP. This was discussed with the ECP coordinator who planned to address this during routine plant department outreach meetings.
During the staff interviews, there was a consistent theme that lower-level CAP issues, even those involving conditions adverse to quality (CAQs), were not addressed timely. There was a perception amongst interviewees that station management didnt see resolution of these issues as a priority, resulting in their being allowed to languish. One example of this was a long-standing open corrective action to replace an incorrectly installed seal purge line on the 3B reactor recirculation pump. This issue, which was documented in IR 590514 as a CAQ, was identified in 2007 and its associated corrective action to re-align the line has been continuously extended, with a current due date of 12/30/2025. Based on licensee evaluation, the misalignment posed a low risk to the pump operation and long-term reliability; however, it was still considered a CAQ as the pump, and by extension the seal purge line, were considered safety-related components. The failure to address this issue for such an extended period lends credence to the staff's continuing view that the station is tolerant towards long-standing issues which, over the long-term, could challenge the site safety culture. The licensee documented this issue as IR 4847173.
No findings or violations were identified.
Failure to Identify a Condition Adverse to Quality with the Local Power Range Monitors Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000237,05000249/2025010-01 Open/Closed
[P.1] -
Identification 71152B The inspectors identified a Green finding and associated Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, when the licensee failed to identify as a Condition Adverse to Quality (CAQ), that the transformers used in the local power range monitor (LPRM) power supplies were of unanalyzed design, and therefore, a potential operability issue affecting the safety-related average and oscillation power range monitoring systems.
Description:
The LPRMs are safety-related components that measure neutron flux at various locations within the reactor core. They feed signals into the average power range monitors (APRMs)which provide control room indication of average reactor power and generate an automatic shutdown of the reactor should reactor power get too high. At least 50 percent of the LPRM signals are required for each APRM to be considered operable. The LPRMs also input into the Oscillation Power Range Monitors (OPRM) which are used to monitor for reactor flux instabilities which could challenge fuel design limits. Like the APRMs, the OPRMs provide control room indication and generate automatic reactor shutdown signals should reactor power conditions challenge the fuel design bases.
In 2017, the licensee identified obsolescence concerns with the LPRM circuit cards and power supplies; in that the original equipment manufacturer (OEM) was no longer manufacturing them or providing support. In response, the licensee decided to have another vendor re-engineer the components to the OEM specifications. Since they would be in effect using the same components, the licensee considered this as an Item Equivalency instead of a Design Change. Subsequently, the licensee replaced the OEM components with the re-engineered, equivalent components. This action was also done at the Quad Cities and LaSalle nuclear power stations, as they utilized the same system and had the same obsolescence issues.
On March 13, 2024, the licensee-initiated Issue Report (IR) 4757787 to document an adverse trend in LPRM failures. Further troubleshooting identified that the issues were related to the re-engineered LPRM circuit cards and power supplies. The affected components were sent to the vendor for failure analysis. On February 6, 2025, the licensee documented that the vendor had identified that the issue was premature failure of the power supply transformer due to increased temperature during operation. However, the IR was closed without any additional detail regarding the cause of the increased temperature nor any CAs.
On March 4, 2025, during subsequent conversations with licensee engineering staff, the inspectors learned that the heating was due to the power supply transformer being of a different design than the OEM. Specifically, the OEM transformer comprised two bobbins with separated primary and secondary coils, while the re-engineered transformer consisted of a single bobbin where the primary and secondary coils overlap. As a result, the re-engineered components were not equivalent and therefore, were of unanalyzed design. The inspectors identified that the premature failure of the LPRMs from overheating, due to the power supply transformers not being equivalent to the OEM, was not identified as a CAQ by the licensee. Therefore, there were no formal CAP actions to evaluate and correct the potential operability/functionality impacts from the unanalyzed design.
The inspectors reviewed the Item Equivalency Evaluation that had been performed for the circuit boards and power supplies prior to their installation and noted that the vendor had documented that due to there being no OEM specifications for the transformer, the vendor had built a replacement having the same physical dimensions and using equivalent functional parameters. It is unknown if the licensee was aware of the differences in construction between the OEM and re-engineered transformer; however, the re-engineered components were accepted and installed as equivalent.
Corrective Actions: The licensee documented the issue in the CAP and was working with the vendor to develop a replacement transformer and assessed that sufficient LPRMs were available to ensure the operability of the APRMs and OPRMs.
Corrective Action References: The licensee documented this issue in the CAP as IRs 4843976 and 4843370. Subsequently, the vendor issued a Part 21 on these power supplies, after identifying several additional failures at Quad Cities and LaSalle. This Part 21 was documented as AR 4848680.
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to identify a CAQ as required by station procedures, was a performance deficiency. Specifically, step 4.1.2 of PI-AA-125, Corrective Action Program (CAP) Procedure, revision 9, states in part that, If at any time a SCAQ or CAQ or any question of either current or past Operability/Reportability arises, then initiate an Issue Report in accordance with PI-AA-120. As stated, the LPRM transformers were not equivalent to the OEM and therefore, were of unanalyzed design. This was a CAQ, as it potentially affected the operability and functionality of a safety-related component. Therefore, an IR should have been initiated in the CAP once the licensee became aware of this design discrepancy.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the LPRMs feed into both the APRMs and OPRMs, both of which have safety functions to monitor reactor power and provide both control room indication and automatic reactor shutdown signals should conditions warrant that could challenge the fuel design limits.
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 C, and answered No to all of the screening questions. Therefore, the finding screens to very low safety significance (Green).
Cross-Cutting Aspect: P.1 - Identification: The organization implements a CAP with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to initiate an IR after identifying that the transformer used in the re-engineered LPRM power supplies was not an equivalent component and therefore posed a potential operability issue as it resulted in an unanalyzed design condition.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures be established to ensure that conditions adverse to quality, such as deficiencies and nonconformances, are promptly identified.
Contrary to the above, between February 6, 2025 and March 10, 2025, the license failed to identify a condition adverse to quality associated with the safety-related LPRMs. Specifically, when the licensee learned through a failure analysis that the transformers used in the replacement power supplies for the safety-related LPRMs were not equivalent to the OEM and that these transformers caused the power supplies to overheat and fail, the licensee failed to enter the non-equivalent design into the CAP until after being questioned by the inspectors on March 4, 2025.
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 March 21, 2025, the inspectors presented the biennial problem identification and resolution inspection results to H. Patel, Plant Manager and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
25499
Results of In-Line Inspection for Unit 3 CCSW Buried Piping
4055891
No Procedural Guidance for ADS Inhibit Failure
22239
U1 DOP 5750-29 Procedure Revision Required
246593
FRAC Tank Contents Sent to Floor Drain Unfiltered
259869-90
Dresden Review of IRIS 471626-MRC 01/21/2020
274841
FME: FM Retrieved from 2D Shutdown Cooling Heat
Exchanger
290588
Dose Rate Alarm While Working in the Radwaste Basement
299223
NDE-UT Results Below Minimum Required Thickness
4365128
Elevated Temperatures on 2/3 EDG Clg Wtr Sys
4381233
Bus 33-1 Cubicle Has High Closing Coil Resistance
4403530
Count Room Laboratory Humidifiers
24280
Obsolescence of GE CR124 Overload Relays
25847-15
CAPE to Design Mech Eng. to Determine Cause of U3 "A"
MSL Inboard MSIV Partially Closed. Address Known Risks
for Other MSIVs and Actions to Prevent Recurrence
26543
Breaker From Bus 30 to Bus 20-3 Would Not Close
4439802
EC for U2 SLMS System Upgrade
4444309-19
Dresden Review of IRIS 502279-MRC 12/13/2021
4453378
U3 Auto Scram due to TR3 Failure and Fire
4454158
U3 EDG Failed To Start
4454768
4.0 Critique-Fire Brigade Response to MPT-3 Fire
4459334
2B MSIV Close Time Was Slow
4459334
2B MSIV Close Time Was Slow
4465446
OPEX Review of IRIS Report 490985
4470594
Missed Opportunity for Thermography Dat Analysis
4471720
RT-3 Trouble
4476471
SBO Bus 61 FD to Bus 24 HCAR-IV Needs to be Replaced
Corrective Action
Documents
4484453
U2 EDG Circ and Turbo Pump Breaker Tripped at MCC 28-3
4487470
Failure Analysis for Siements MPT 3 H3 Bushing Not
Performed
4494317-03
Engineering to Complete SA for CRDs
4497371
Trend IR: DC System Grounds
4504957
Evaluate Degraded U3 125VDC Battery Cables
4504959
Evaluate Degraded U3 125VDC ALT Battery Cables
4504995
Evaluate Degraded U3 250VDC Battery Cables
4513269
AO 3-1601-24 Valve Stem Sheared While Performing
DIS 1600-24
4513269
AO 3-1601-24 Valve Stem Sheared While Performed
DIS 1600-24
4513674
2/3 Lift Station Battery Performance Test Results
4513810
AO 3-1601-62 Failed to Open
4513848
Bar Rack Plugging DOA Entries
4513848
Bar Rack Plugging DOA Entries
4514409
Licensee Renewal Implementation Program Routine Taks #4
4533061
MOV 3-1501-20B As-Found Unsat Criteria for Low CST
Thrust
4534332-16
Dresden Review of IRIS 542867 01/17/23
4534332-18
Dresden Review of IRIS 540504 01/24/23
4534332-22
Dresen Review of IRIS 543604 2/7/23
4534332-36
Dresden Review of IRIS 566271-MRC 09/07/23
4534332-39
Dresden Review of IRIS 54644 03/14/23
4534332-65
Dresden Review of IRIS 560600 06/26/23
4535914
Cranes Handling Fuel Need to Have Manual Lowering
Means
4539818
Unexpected Load Drop
4539877
2A LP Turbine Steam Sealing Line Leak (update to IR 4509999
4540536
MRULE A1 Determinations Performed Incorrectly
4544622
Cables to Bus 37 Have an Acrid Odor
4545056
Steam Leak Identified in 2B SJAE Room
4550338
Corrective Action
Documents
4550338
NRC ID WO Lacked Evaluation for LLRT [Local Leak Rate
Test] PMT [Post Maintenance Test] of MSIV [Main Steam
Isolation Valve] Packing Replaced
4551420
2/3 EDG Feed to Bus 33-1 Breaker Cubicle Racking Issue
4552509
Cyber - NRC Observation - CC-AA-606 Scope
4553868
LORT DEP Not Included in PI Submissions
4555262
Trend IR: Burnishing Relay Contacts
45557716
Elevated Iron and Copper in U2/3 EDGJCW System
4555864
IC Pumphouse TIC Not Procured SQAD-7 Requirements
4556532
Operating Experience Applicable for Recirculation Oil
Leakage-Managing Risks
4557966-01
CC-AA-601 : assignment list 2023
4557966-02
List of PCRA for CC-AA-601
4558782
NOS ID: WHR Issues in Maintenance
4560811
Clinton NRC Cyber Security Inspection Issues
4560825
SPC 04461720-30 Failure Analysis of MPT 3 Thermoplate
Result
4560825
SPC 04461720-30 Failure of MPT 3 Thermoplate Result
4561068
Trend IR: U2 RBCCW [Reactor Building Component Cooling
Water] High Leakage Rate
4561990
3-0203-3A PS Controller OOT
4562297
Transducers Mounted in 903-29 Panel Without Seismic
Review
4562451
NRC CS - Firewall Tuning Not Documented
4562593
Unusual External Force on Pump-3C Condensate Pump
4563898
Unexpected Response during Dis 287-01 on D2
4563898
Unexpected Response during DIS 287-01 on D2
4564021
Receipt of NRC PI&R Insp Rpt 2023010 - Finding
23010-01
4564026
Receipt of NRC PI&R Inspection Report 2023010-Violation
23010-03
4564037
Receipt of NRC PI&R Inspection Report 2023010-Violation
23010-03
4564041
Receipt of NRC Inspection Report 2023010-Vilation
23010-04
4666025
23 NRC Graded Exercise TSC Performance
4670279
LMS Inaccuracies Regarding NRC RO and SRO License
Dates
4671637
OPXR to Maintenance Planning to Review the OE in IRIS
549630
4671637-03
OPXR to Chemistry to Review the OE in IRIS 549630 and
Verify No Similar Issues Exist with their Dynamic Procedures
4672099
MPT-3 Degraded Cables
4672164
D3M21 Chemistry Samples Showing Elevated Dose
4681273
Unexpected Alarm 903-6 H-9, 3C RFP Brg Oil Press Lo
4681519
U3 Reactor Building Fire Protection Supply Header Leak
4681955
Review of Station Events Requiring IRIS Reporting
4681967
NRC 1Q23 Inspection Report 2023001
4683443
Train Alignments of 3C CD/CB Pumps
4684857
3-1599-80C Broken Hand Wheel
4685104
3C CCSW Pump Diff Press at High End of Acceptable
Range
4686063
Trend IR for Past Late PMs in Maintenance
4686063
Trend IR for Past Late PMs in Maintenance
4688524
3A CCSW Pump Inboard Seal PI 3-1541-5A Possibly
Plugged
4688989
U2 125VDC Alt Batter Feed Ckt Breakers, SR Components
Req'd
4688989-07
Perform CAPE on U2 125 VDC Alt Batt Feed Ckt Bkrs, SR
Comp
4690019
NRC ID : Documentation for DOS 6600-08 Incomplete
4691476
NRC Identified: Frisker Probe Marked Broken
4691637
23 CETI: ASME Code Related Component ISI/ANI
Documentation
4692850
U2 Target Rock (2-0203-3A) Temperature Increase
4694214
HPCI Signal Converter Servo Amp Hardening Opportunity
4695296
23 CETI-NRC ID: Issue with Uncertainty Not in ECCS
[Emergency Core Cooling System] Testing
4695924
23 Guided Wave U3 CCSW Inspection Results
4700056
3-0203-3B Pressure Controller Switch Out of Tolerence
Corrective Action
Documents
4700263
Issues Attempting to Perform SAWA Performance Test
4700513
Dresden Review of IRIS 563905-MRC 9/11/23
4700513
Dresden Review of IRIS 566713-MRC 09/07/23
4700513-20
Dresden Review of IRIS 569732-MRC 10/10/23
4700948
Part 21 GE Hitachi Ultra High Duty and Medium Duty Control
Rods
4701702
Perform Level 3 OPEX Review of WANO SER-2023-02
Complete Form
4701772
2/3 RB Breaker (24-1 cube 1) Will Not Rack Full In
4702234
Temporary Patch Leak on Fire Piping Downstream of
2/3-4199-176
4702817
Receipt of NRC CETI [Comprehensive Engineering Team
Inspection] Inspection Report 2023011-Violation 2023011-01
4702820
Receipt of NRC CETI Inspection Report 2023011-Violation
23011-03
4703442
3D CCSW Pump Inboard Seal Leak
4705130
U3 EDG Fire Door Inoperable
4711018
U2 RPS EHC CV Pressure Switches Hydraulic Braided
Hoses
4713088
MOV 2-1501-32A Degraded Grease
4714001
APS ID'd Valve Assembly Installed in Wrong Location
4714653
Rebuild ERV Pilot Valve 2-0203-3D-PIVL
4715155
287-106B Relay Required Burnishing
4716981
Unexpected Alarm: 902-4, D-23 2A Target Rock Relief Valve
INOP
4717468
Transfer Time of MCC 28-7/29-7 From Bus 29 to Bus 28
4717499
U2 Div 2 CCSW Supply to SW Flow Low During DOS
1500-12
4717529
CCSW Support M-1164D-129 Post VT3 Exam Recordable
Indication
4718350
D2R28 LL: Unexp High Therm Limits During U2 Power
Ascension
4718861
Air Void Identified U2 HPCI [High Pressure Core Injection]
during 1M Periodic Verification
22035
Biocide Chemical Leak on EDG Diesel Piping
23409
Security - (2) Perimeter Term Boxes Failed Tamper Testing
27126
4kV Circuit Breaker Trend IR for Last 6 Months of 2023
4730727
SPC 04696436-03 Failure Analysis of U2 EDG Control
Switch
4733144
OPEX Review-Gasket Selection
4736846
U3 CCSW Vault Test Door Warped
4740605
3Q2023 NRC Report-NCV 05000237,05000249/2023003-01
4740607
3Q2023 NRC Inspect Rpt-NCV 05000237,05000249/2023003-02
4741885
Offgas Release Rate Errors Identified (APEX software)
4750761
Wiring Discrepancy Identified during Verification
4750971
U3 CCSW Pump Vault Flood Protection Leakage Test
Results
4751884
TRNG ID : Level 3 OPEX Review
4753182
U3 HPCI GLSO MCC Wiring Degradation
4753182
U3 HPCI GLSO MCC Wiring Degradation (2/26/24)
4754250
U3 HPCI Aux Cooling Water Pump
4757240
D2 Emergency Diesel Generator Resistor Mounting
4758041
Unit 3 Drywell Pressure Response
4758041-15
Perform CAPE on 04758041 Unit 3 DW Pressure
4760311
OPEX Review--Sequoyah NRC Inspection Report and
Finding
4760717
Unit 3 Feedwater Iron in Action Level 1
4761885
Perform Site OPEX Level 3 Evaluation for IRIS 575027
4761938-04
NRC Functional Engineering Inspection Aging SSCs
71111.21N.04
4762585
U3 CCSW Vault Penetration High Leakage
4763657
Resolution To Air Void in U2 HPCI Discharge
4763825
U3 Instrument Air Compressor a(1) Action Plan
4764162
3-1501-1A Has Back Leakage to Bay 13
4769600
U3 Battery Charger Load Bank Faulted during Testing
4771001
Buzzing Sound Heard from U2 125 VDC Alt Charger
4771800
3A SDC Pump Breaker Closing Coil Continuity High
4773245
24 NOS CAP Audit
4775724
3.3.e Required Actions Not Met
4776882
Preparation for NRC PI&R Inspection
4780148
U2 CCSW Vault Test Door Cracked Welds
4780342
2-0203-3B ERV Tailpipe Temperature Rising
4781967
Leak Downstream of Discharge Flange on 3A Closed
Cooling Service Water Pump
4783211
4785519
IER L4 24-1 Human Performance Errors Leading to Events
47880262-04
Dresden Review of IRIS 590375-MRC 07/10/24
4788577
EO Rounds: Load-Bearing Scaffold Pole on LPCI/CS Piping
4789300
EP Exercise Critique Comments
4789511
U2 Discharge Canal Temperature Recorder-EPN: 2-4450-66
4790201
Met Tower to MCR Equipment Going Obsolete
4790409
ERV Stainless Steel BOA Found in Degraded Condition
4791758
Breaker Failed to Stay Shut
4792614
IRIS 596925 Degrading Condenser Vacuum - Site OPEX
Reviews
4792614-09
Level 3 OPEX Evaluation : IRIS 596925 - Prairie Island Unit
2, Degrading Condenser Vacuum Results in Power Reductio
4796772
Potential Trend IR: Unit 2 Drywell CAM Equipment Failure
4797011
2Q24 NRC Integrated Inspection Report-NCV 2024002-01
4797051
2Q24 NRC Integrated Inspection Report-NCV 2024002-02
4797062
NRC Integrated Inspection Report-NCV 2024002-03
4797062
3Q2024 NRC Integrated Inspection Report-NCV 2024002-03
4797345
Water Leaking onto Cable Tray in U3 Moisture Seperator
Area
4797345
Water Leaking onto Cable Tray in U3 Cable Seperator Area
4799921
Erratic Offgas Flow during Unit 2 Downpower
4799972
Vacuum Leakage on 2-3103-B3 Heater at Reduced Power
4800029
Leakage Identified on 2-3099-113A
4803442
2/3 Diesel Fire Pump Failed Overcrank Test
4804468
Elevated Conductivity for Unit 3 Condensate Pump
Discharge
4806177
EPID: ERO Min Staff Individual Lives Outside Response
Time
4806275
Receipt of NRC Finding/NCV : Vital Area 31 Day Access
Review
4807695
ECP Requires Revision
4807773
48113364
Bus 33-1 Loads Did Not Load Shed During UV (DOS 6600-
03)
4814260
QDC SDV LLRT OPEX Applicable to Dresden
4814659
3Q2024 NRC Inspection Report: NCV 05000237/2024003-01
4815394
Abbreviated Maintenance 3A Target Rock Sensing Line
Fitting
4815440
D3R28 Drywell Aging Management Program A.1.26
4816220
TRM Entry for Fire Barrier Impaired
4817063
Bus 34-1 Feed to Bus 39 Breaker Won't Charge
4817876
U3 CRD C-09(10-35) Triple Notched
4818393
Drywell Inlet/Outlet Isolation Valves
21853
2B CD/CB Motor Vibration SME Call Recommendations
23934
2A CS Closing Coil Continuity High
24753
NRC Identified Issue
24762
IR 04817965 Historic Operability Review
24921-03-00
MRULE a(1) Determinations Not Performed
28862
Evaluation of Bar Rack Removal for Frazil Ice Mitigation
4832856
3A CCSW Pump Motor Trend - Elevated Amps
4833604
PI&R SA ID: DRE Aging Management ER-AA-700 Gaps
4834985
Fire Header Leak
4837255
NRC RIS 2025-02
4842823
U2 Condenser In-Leakage at Main Condenser Pipe
590594
3B Recirculation Seal Purge Line Piped to Wrong
Connection
Corrective Action
Documents
IA Dryer 'A' Tower
Blowing Down
Excessively
3B
4843370
NRC ID: LPRM-Paragon ICPS Failure Analysis Report
4843391
NRC ID: LPRM Ion Chamber Power Supply Testing
Corrective Action
Documents
Resulting from
4843976
NRC ID: IR to Capture ICPS Transformer Failure
4845429
NRC ID: OPXR ATI 4780262 -63 Closure
4845454
NRC ID: Comments Related to IR 4564037
4846007
NRC ID: Enhancement Opp Regarding OPEX Review
Actions
4846606
Incorrect Maintenance Rule Function Screened for
4846642
NRC/IEMA Walkdown U3 250VDC Battery Cable Bend
4846643
NRC/IEMA Walkdown U2 15VDC Battery Cable bend/crack
4846647
NRC/IEMA Walkdown U2 250VDC Battery Cable Bend
4846746
NRC ID: OPXR review improper completion
4846807
NRC ID PI&R Comments Regarding Trend IR# 4727126
4846830
MRule Risk Significance Function15-6 (NRC ID -PI&R 2025)
4846837
NRC ID - Engage Health Severity Level Miss-Categorized
4846898
NRC Identified (PI&R 2025): U3 125VDC Battery Cable
Crack
4846901
NRC Identified (PI&R 2025): Cable bend evaluation U3 125
VDC
4846972
NRC Identified (PI&R 2025): Initiate Trend of DC Grounds
Inspection
4847173
NRC ID: Low Safety Significant Issue Allowed to Linger
U3 125VDC System Battery Ground
Engineering
Changes
ECR 461979
U3 HPCI GLSO Bucket Wiring Degradation
Nuclear Safety Culture Meeting Minutes for Monitoring
Period 1/1/2024-4/30/2024
05/20/2024
Nuclear Safety Culture Review Committee Meeting Minutes
for Monitoring Period 9/1/2024-12/31/2024
01/27/2025
Plant Health Committee Meeting Agenda and Presentation
Materials
2/24/2025
Station Oversight Meeting Package and Minutes
03/05/2025
Management Review Committee Package and Minutes
03/06/2025
Employee Concerns and Safety Culture Monitoring Program
Monthly Report
2/13/2025
0101-0099-RPT-
001
Dresden Clean Energy Center Evaluation of Mitigations to
Intake Grassing Events
Miscellaneous
DR-PBD-AMP-
XI.M20
Open-cycle Cooling Water System GALL-SLR
Revision 0
Management
Review
Committee
Agenda
03/05/2025
PHC 2.10.2025
Plant Health Committee Meeting per ER-AA-2001
2/10/2025
Station Oversight
Committee
Agenda
03/04/2025
Operability
Evaluations
44619342C MSIV
1B Limit Switch
Found Failed
DMP 0200-15
Main Steam Isolation Valve Maintenance
Revision 39
DMP 0200-43
Main Steam Isolation Valve Operator Air and Oil Cylinder
Overhaul
Revision 10
DMP 0200-44
Main Steam Isolation Valve Operator Air Manifold
Replacement and Air Operator Testing
Revision 03
DOP 4400-07
Circulation Water De-Icing Operation
Revision 15
Employee Concerns Program
Revision 11
Employee Concerns Program
Revision 15
ER-700-1003
Use of Operating Experience for License Renewal
Implementation/Aging Management
Revision 6
Maintenance Rule Implementation Per NEI 18-10
Revision 0
Maintenance Rule (a)(1) and (a)(2) Requirements
Revision 3
License Renewal Implementation Program
Revision 10
Station Rework Reduction Program
Revision 14
MA-AA-716-230-
1003
Thermography Program Guide
MA-DR-EM-3-
83044
25 VDC Battery Performance Test
Quality Assurance Topical Report
Revision 99
OP-AA-108-105-
1001
MCR and RWCR Equipment Deficiency Management and
Performance Indicator Screening
Safety Culture Monitoring
Revision 4
Operating Experience Program
Revision 7
Operating Experience Program
Revision 7
Procedures
Processing of Level 3 OPEX Evaluation
Revision 7
Issue Identification and Screening Process
Revision 13
Corrective Action Program (CAP) Procedure
Revision 9
Effectiveness Review Manual
Collective Effectiveness Review
Nuclear Safety Culture
Revision 0
Controls for Radiation Protetion Instrumentation
Revision 10
On-Line Scheduling Process
Preventative Maintenance (PM) Database Revision
Requirements
Revision 6
Security NOS ISFSI 2024 Audit
04/23/2024
NRC 71130.09 Security Plan Changes
2/12/2024
Security NOSA 2025 Audit
2/17/2025
01307278-68
Dresden Configuration Control Program
04388146-76
Clearance and Tagging
04556064-02
Security Pre-NRC Self-Assessment for Fitness for Duty
2466136-91
Maintenance Rule Periodic Assessment#15 (10CFR50.65
(a)(3) Assessment)
2/19/2024
4761938-04
Self-Assessment: Age-Degradation Self-Assessment--GAPs
Identified
4763024-04
IER L1-17-5 Revision 1 : Line of Sight to the Reactor Core
4777143
Self-Assessment 2024 Nuclear Oversight Employee
Concerns Program
NOSA-DRE-24-
Engineering Design Control Audit Report
Self-Assessments
NOSA-DRE-24-
Corrective Action Program Audit Report
01742198
Transfer Spare Main Power Transformer from Quad Cities
206323-23
26226
D3 10Y PM 250VDC Bkr HPCI Turb Gland Seal Cond
Hotwell Drain
29409
SBO Bus 61 FD to Bus 24 HCAR-IV Needs to be Replaced
238793
U2 EDG Circ and Turbo Pump Breaker Tripped at MCC 28-3
257567
2B MSIV PM Electromagnetic Relief 'D' Replace Pilot
08/02/2023
Work Orders
289342-01
OP Perform As-Found Leak Test Per DOS 1500-21
5332613
Evaluate Degraded U3 125VDC Battery Cables
5598013
Bus 34-1 Feed to Bus 39 Breaker Won't Charge