IR 05000298/2025003
| ML25345A166 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/18/2025 |
| From: | Douglas Dodson NRC/RGN-IV/DORS/PBC |
| To: | Dia K Nebraska Public Power District (NPPD) |
| References | |
| IR 2025003 | |
| Download: ML25345A166 (0) | |
Text
December 18, 2025
SUBJECT:
COOPER NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000298/2025003
Dear Khalil Dia:
On September 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Cooper Nuclear Station. On November 20, 2025, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.
Due to the temporary cessation of government operations, which commenced on October 1, 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 inspection reports that should have been issued by November 13, 2025, to December 31, 2025. The NRC resumed the routine cycle of issuing inspection reports on November 13, 2025.
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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Cooper Nuclear 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 IV; and the NRC Resident Inspector at Cooper Nuclear 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, Douglas E. Dodson II, Chief Reactor Projects Branch C Division of Operating Reactor Safety Docket No. 05000298 License No. DPR-46
Enclosure:
As stated
Inspection Report
Docket Number:
05000298
License Number:
Report Number:
Enterprise Identifier:
I-2025-003-0006
Licensee:
Nebraska Public Power District
Facility:
Cooper Nuclear Station
Location:
Brownville, NE
Inspection Dates:
July 1, 2025, to September 30, 2025
Inspectors:
G. Birkemeier, Resident Inspector
G. Kolcum, Senior Resident Inspector
Approved By:
Douglas E. Dodson II, Chief
Reactor Projects Branch C
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Cooper Nuclear 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 Properly Perform Maintenance Affecting the Performance of Safety-related Equipment Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000298/2025003-01 Open/Closed
[H.12] - Avoid Complacency 71111.15 The inspectors reviewed a self-revealing finding of very low safety significance (Green) and an associated non-cited violation of Technical Specifications 5.4.1.a, "Instructions,
Procedures, and Drawings," for the licensee's failure to properly pre-plan and perform maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly verify tolerances associated with the fit-up of the emergency diesel generator 1 jacket water pump impeller, resulting in its failure and inability to perform its safety function.
Failure to Identify and Correct Conditions Adverse to Quality Associated with External Flooding Protection for the Service Water Pump Room Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000298/2025003-02 Open/Closed None (NPP)71152A The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which requires, in part, that measures shall 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. Specifically, since July 1985, the licensee failed to identify and correct conditions adverse to quality associated with external flooding hazard protection for the service water pump room, as evidenced by: (1) an inadequate evaluation for a procedure change in July 1985; (2) a failure to identify an adverse condition during flooding walkdowns in 2012; and (3) a failure to identify an adverse condition in the Flooding Hazard Reevaluation Review in 2015, which resulted in service water pump room components being susceptible to design-basis flood conditions and a loss of capability to perform their safety functions for their 30-day mission time.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000298/2025-001-00 Inoperable Diesel Generator Jacket Water Pump Results in Condition Prohibited by Technical Specifications 71153 Closed
PLANT STATUS
Cooper Nuclear Station began the inspection period at rated thermal power. On August 8, 2025, the unit lowered power to 65 percent for planned main turbine valve testing and rod sequence exchange. The unit returned to full power on August 9, 2025. The unit remained at rated thermal power for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - 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)250/125 Vdc batteries to inspect battery terminals following operating experience of aging related cracking on July 29, 2025
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the reactor protection system electrical distribution on September 12, 2025.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
(1)control building emergency condensate storage tank room on August 26, 2025 (2)service water pump room and intake structure on September 4, 2025
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the on-site fire brigade training and performance during an unannounced fire drill on September 11, 2025.
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated external flooding mitigation protections in the service water pump room on September 29, 2025.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
(1)reactor equipment cooling heat exchanger maintenance windows (A week 6, B week 7) on July 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) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during rod pattern adjustment on August 8, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated an operations training simulator fire drill with anticipated transient without scram conditions scenario on August 21, 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)diesel generator 1 jacket water pump maintenance on September 17, 2025
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (1 Sample)
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)planned Yellow online risk window for division 1 core spray maintenance on August 5, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (2 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)reactor core isolation cooling keep fill high suction pressure alarm on September 1, 2025 (2)standby liquid control at minimum temperature on September 5, 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)cabling modification associated with the emergency station service transformer modification resulting in substantial reduction in station fire risk on July 24, 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) (2 Samples)
(1)fire pump E maintenance on September 3, 2025
- (2) fire pump D maintenance on September 11, 2025
Surveillance Testing (IP Section 03.01) (2 Samples)
(1)scram discharge volume drain valve time testing on August 21, 2025
- (2) high-pressure coolant injection quarterly surveillance on September 17, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)reactor core isolation cooling quarterly surveillance on September 5, 2025
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
The inspectors evaluated:
- (1) operations training simulator scenario on September 18,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified the licensee performance indicator submittals listed below:
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===
- (1) July 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)diesel generator 1 jacket water pump maintenance on September 16, 2025 (2)follow-up on corrective actions from historical flooding relative to service water design-basis on September 19, 2025
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 05000298/2025-001-00, "Inoperable Diesel Generator Jacket Water Pump Results in Condition Prohibited by Technical Specifications" (ML25157A104). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71111.15. This LER is Closed.
INSPECTION RESULTS
Failure to Properly Perform Maintenance Affecting the Performance of Safety-related Equipment Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000298/2025003-01 Open/Closed
[H.12] - Avoid Complacency 71111.15 The inspectors reviewed a self-revealing finding of very low safety significance (Green) and an associated non-cited violation of Technical Specifications 5.4.1.a, "Instructions, Procedures, and Drawings," for the licensee's failure to properly pre-plan and perform maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly verify tolerances associated with the fit-up of the emergency diesel generator 1 jacket water pump impeller, resulting in its failure and inability to perform its safety function.
Description:
On April 8, 2025, the licensee performed a monthly emergency diesel generator 1 run in accordance with procedure 6.1DG.101, Diesel Generator 31-day Operability Test (IST) (DIV 1), Revision 97. Approximately 15 minutes into the run, the jacket water high-temperature alarm was received. Upon investigation, operators found that jacket water pump discharge pressure was indicating 5 PSIG, which is well below the acceptable value of greater than 20 PSIG. Operations subsequently secured the diesel and terminated the surveillance. During disassembly of the jacket water pump, maintenance discovered the pump impeller had separated from the shaft and was freely rotating. Maintenance personnel also observed damage to the bore of the impeller. The licensee repaired the impeller bore and re-assembled the pump on April 11, 2025, and emergency diesel generator 1 was declared operable following a successful surveillance test on April 12, 2025.
The pump had previously been rebuilt on February 5, 2025, to replace mechanical seals following observed seal leakage during a monthly surveillance for emergency diesel generator 1. Work Order (WO) 5487972, which was used to rebuild the pump in February 2025, provided written instructions to maintenance personnel for the disassembly, repair, and reassembly of the jacket water pump. The WO instructions failed to:
- (1) validate measurement of the stainless steel shaft diameter and bronze impeller bore diameter to account for changes in protrusion depth over previous rebuilds (1986, 1993, 2014, and 2024),to ensure tolerance fit (0.0017, as stated as a note in the WO) of impeller seating to the shaft, and to ensure the tolerances of the individual parts interact such that the collective effect is within acceptable limits when combined as a pressed assembly;
- (2) incorporate a tolerance for impeller protrusion dimension of 1/8; and
- (3) perform quality-controlled validation of critical impeller protrusion. The licensee recorded the as-found and as-left impeller protrusion as 0.120, which did not meet the acceptance criteria specified in step 32 of the WO (1/8). Following the pump reassembly, and prior to the failure on April 8, 2025, emergency diesel generator 1 successfully passed its monthly surveillance tests on February 5, 2025, and March 3, 2025, with approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of run time during each test.
Corrective Actions: The licensee implemented corrective actions to revise jacket water pump assembly instructions to validate impeller bore critical dimensions and shaft placement in accordance with vendor guidance.
Corrective Action References: This issue was entered into the licensee's corrective action program as Condition Report CR-CNS-2025-01839.
Performance Assessment:
Performance Deficiency: The licensee failed to pre-plan and perform maintenance that can affect the performance of safety-related equipment in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances, as required by Technical Specification (TS) 5.4.1a. Specifically, on February 5, 2025, the licensee failed to validate impeller bore critical dimensions and shaft placement in accordance with vendor guidance during reassembly of the emergency diesel generator 1 jacket water pump.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality 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 licensee failed to ensure that procedures affecting the performance of safety-related equipment incorporated vendor requirements, which adversely impacted the ability of the component to perform its safety function.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, "The Significance Determination Process for Findings At-Power." The inspectors screened the finding using Exhibit 2, "Mitigating Systems Screening Questions." The inspectors determined that the finding required a detailed risk evaluation to determine significance because the finding is not a deficiency affecting the design or qualification of a mitigating SSC, does not represent a loss of the probabilistic risk assessment (PRA) function of a single train TS system, but does represent a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time.
A senior reactor analyst performed a detailed risk evaluation using the Cooper SPAR model, version 8.82, ran on SAPHIRE, version 8.2.12, to determine that the finding was of very low safety significance (Green). The analyst modeled the finding as a failure to run by the diesel generator. In applying the model, the analyst assumed that the finding had a 62-day T-period (where the failure could occur) and a 3.5-day repair time, which combined for a 65.5-day exposure time. The following model modifications were made to more realistically reflect the configuration and operation of the plant:
- (1) modification of the model to credit the high-pressure coolant injection as an injection source, which could be supported by FLEX equipment in case the reactor core isolation system failed during an event;
- (2) crediting the 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of successful run time after the degradation was introduced;
- (3) adjustment of station battery depletion time to 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to reflect operators shedding loads from the batteries to extend the time available to implement FLEX strategies;
- (4) crediting a plant modification in the licensees fire model, which prevented fire scenarios from impacting the emergency station service transformer; and
- (5) eliminating invalid equipment test and maintenance combinations prohibited by plant procedures during the repair time. In applying the assumptions and modifications, the analyst estimated the increase in core damage frequency from internal events to be 3.4E-7/year. For external events only, fire was a significant contributor for increasing core damage frequency and was estimated to be 5.0E-7/year, leading to a total increase in core damage frequency of 8.4E-7/year (Green). In estimating the increase in large early release frequency, the analyst used IMC 0609 Appendix H, Containment Integrity Significance Determination Process, to apply a factor of 1.0 to the high-pressure core damage frequency sequences to estimate the increase in large early release frequency of 5.8E-8/year (Green). The dominant core damage sequences were station blackouts mitigated by the sites diesel generators and FLEX equipment.
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 failed to perform a thorough review to ensure work procedures incorporated vendor requirements and relied on past successes, which used inadequate work instructions to perform critical steps.
Enforcement:
Violation: Technical Specification 5.4.1.a, requires, in part, that procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 9 of Appendix A to Regulatory Guide 1.33, Revision 2, requires that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.
Contrary to the above, on February 5, 2025, the licensee failed to pre-plan and perform maintenance that can affect the performance of safety-related equipment in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.
Specifically, WO 5487972, which provided work instructions for rebuilding the safety-related emergency diesel generator 1 jacket water pump, was not properly pre-planned and failed to provide sufficient guidance for maintenance personnel to perform verification of tolerances associated with the fit-up of the pump impeller to the shaft in accordance with design drawings.
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 Conditions Adverse to Quality Associated with External Flooding Protection for the Service Water Pump Room Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000298/2025003-02 Open/Closed None (NPP)71152A The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which requires, in part, that measures shall 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. Specifically, since July 1985, the licensee failed to identify and correct conditions adverse to quality associated with external flooding hazard protection for the service water pump room, as evidenced by:
- (1) an inadequate evaluation for a procedure change in July 1985;
- (2) a failure to identify an adverse condition during flooding walkdowns in 2012; and
- (3) a failure to identify an adverse condition in the Flooding Hazard Reevaluation Review in 2015, which resulted in service water pump room components being susceptible to design-basis flood conditions and a loss of capability to perform their safety functions for their 30-day mission time.
Description:
By letter dated March 12, 2012, the U.S. Nuclear Regulatory Commission (NRC)issued a request for information to all power reactor licensees and holders of construction permits in active or deferred status, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.54(f), "Conditions of license" (hereafter referred to as the "50.54(f)letter"). The request was issued in connection with implementing lessons-learned from the 2011 accident at the Fukushima Dai-ichi nuclear power plant, as documented in The Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident.
Recommendation 2.1 in that document recommended the NRC staff issue orders to all licensees to reevaluate seismic and flooding hazards for their sites against current NRC requirements and guidance. Subsequent staff requirements memoranda associated with Commission Papers SECY-11-0124 and SECY-11-0137 instructed the NRC staff to issue requests for information to licensees pursuant to 10 CFR 50.54(f).
By letter dated February 3, 2015, Nebraska Public Power District (NPPD, the licensee)submitted its Flood Hazard Reevaluation Reports (FHRRs) for Cooper Nuclear Station (Cooper) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15041A523).
The inspectors noted that on July 1985, the licensee established procedure 5.1.3, Flooding, and removed the requirement to barricade the doors to the service water pump room, which is located in the intake structure. The licensee developed new procedure 5.1Flood, Flood, which replaced procedure 5.1.3, Flooding, and the new procedure no longer had any actions to protect the service water pump room to withstand a design-basis flood without loss of SSC capability to perform their safety functions for their 30-day mission time.
The inspectors also noted that during flooding walkdowns in 2012 for the post Fukushima flooding walkdowns, the licensee failed to identify that the personnel access commercial steel doors with louvers into the service water pump room provided no water tightness or resistance to static or dynamic forces during the design-basis flood into the service water pump room. In addition, susceptible components like the backwash solenoid operated valves for the strainer and the inlet to the service water pump motors were below the newly re-evaluated design-basis flood level of 908.37 feet.
Finally, the inspectors noted that during 2015, the licensee submitted the FHRRs and stated, Wave action at the Intake Structure will not affect the safe shutdown of the plant since the service water pumps and controls are protected by massive reinforced concrete walls and slab up to elevation 919-0. This is also stated in chapter II, Section 4.2.2.2, Site Flooding Protection, of the Updated Final Safety Analysis Report (UFSAR). In addition, the FHRR states, The Intake structure need not be isolated due to the fact that there is no essential equipment located there at or below 906 feet MSL that will be adversely affected in performance of a safe shutdown function in the event of flooding. In 2025, the Resident Inspectors conducted walkdowns of the service water pump house and associated flooding protection measures. The inspectors identified components critical to the performance of safe shutdown equipment located below the previously identified threshold of 906 feet MSL. The licensee conducted further inspection to verify components below the protected height.
Specifically, solenoid valves required to conduct strainer backwash flow were found below this level. The service water pumps require strainers and strainer backwash flow to remain operable. The backwash solenoid operated valves for each train are approximately 22 inches and 30 inches from the floor (903-6) for train A and train B, respectively, making the electrical components susceptible to failure during an external design-basis flooding event.
Corrective Actions: The licensee placed the issues into the corrective action program.
Corrective Action References: This issue was entered into the licensees corrective action program as Condition Report CR-CNS-2025-01623.
Performance Assessment:
Performance Deficiency: The licensee failed to adequately identify, evaluate, and correct conditions adverse to quality in accordance with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the service water pump room. Specifically, since July 1985, the licensee failed to correct conditions adverse to quality associated with external flood hazard protection for the service water pump room, as evidenced by:
- (1) an inadequate evaluation for a procedure change in July 1985;
- (2) a failure to identify an adverse condition during flooding walkdowns in 2012; and
- (3) a failure to identify an adverse condition in the Flooding Hazard Reevaluation Review in 2015, which compromised the ability of the service water pump room components to withstand a design-basis flood and perform their safety function for the 30-day mission time.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors 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 licensee failed to ensure adequate flood protection existed to protect SSCs in the service water pump room, which resulted in service water pump room components being susceptible to design-basis flood conditions and a loss of capability to perform their safety functions for their 30-day mission time.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Specifically, using Exhibit 2-1, Section B, of IMC 0609, the inspectors determined that a detailed risk evaluation was required for this finding because the finding involved the loss or degradation of equipment designed to mitigate flooding for greater than 14 days. A senior reactor analyst performed a detailed risk evaluation using the Cooper SPAR model, version 8.82, ran on SAPHIRE, version 8.2.12, to estimate that the finding was of very low safety significance (Green). In performing the detailed risk evaluation, the analyst assumed the flood frequency that would render the service water system nonfunctional was 1.0E-6/year, based on information provided from individuals in the Engineering and External Hazards division of the Office of Nuclear Reactor Regulation. The analyst assumed the effects on the plant from this flooding event were best modeled as a transient with that flood frequency and all service water and all FLEX equipment unavailable. These assumptions yielded an estimate in the increase of core damage frequency of 2.3E-9/year. No further quantification of the increase in core damage frequency was necessary because only risk from external flooding was applicable to the finding. Estimation of the increase in large early release frequency was not performed due to the very low increase in core damage frequency.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance's are promptly identified and corrected.
Contrary to the above, since July 1985, measures were not established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances were promptly identified and corrected. Specifically, since July 1985, the licensee failed to correct conditions adverse to quality associated with external flooding hazard protection for the service water pump room, as evidenced by:
- (1) an inadequate evaluation for a procedure change in July 1985;
- (2) a failure to identify an adverse condition during flooding walkdowns in 2012; and
- (3) a failure to identify an adverse condition in the Flooding Hazard Reevaluation Review in 2015, which resulted in service water pump room components being susceptible to design-basis flood conditions and a loss of capability to perform their safety functions for their 30-day mission time.
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 November 20, 2025, the inspectors presented the integrated inspection results to Khalil Dia, 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
CR-CNS-
20-02133
Drawings
3010
Burns and Roe
Miscellaneous
Reactor Protection System - MG Sets
Miscellaneous
914E175
Electrical Protection Assembly
Miscellaneous
CNS No. 633
Vendor Manual
Miscellaneous
CNS No. 97
Vendor Manual
Procedures
2.2.22
Vital Instrument Power System
Procedures
2.2A
20CRIT.DIV1
20/240 VAC Critical Instrument Power Checklist (Div 1)
Procedures
2.2A
20CRIT.DIV1
20/240 VAC Critical Instrument Power Checklist (Div 1)
Procedures
2.2A
20VITAL.DIV3
20/240 VAC Vital Instrument Power Checklist
Procedures
2.2A_125DC.DIV1
25 VDC Power Checklist (Div 1)
Procedures
2.2A_250DC.DIV1
250 VDC Power Checklist (Div 1)
Procedures
6.1RPS.313
RPS Electrical Protection Assemblies Calibration and
Functional Test (Div 1)
Procedures
6.2RPS.313
RPS Electrical Protection Assemblies Calibration and
Functional Test (Div 2)
Drawings
CNS-FP-256
Intake Structure
Drawings
Fire Protection Pump House
Drawings
Off-Gas Building
Drawings
Transformer Yard
Miscellaneous
NEDC 11-089
Fire Safety Analysis for Fire Area IS-A
Procedures
0-BARRIER-
MAPS
Barrier Maps
Procedures
5.4POST-FIRE-
REACTOR
Reactor Building Post-Fire Operational Information
Procedures
TPP 207
Fire Brigade Drills and Evaluations
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
NEDC 91-221
Service Water Pump Room Temperatures After Loss of
Cooling
Procedures
NEDC 09-102
Internal Flooding - HELB, MELB, and Feedwater Line Break
Corrective Action
Documents
CR-CNS-
25-01726, 2025-01744
Procedures
2.0.2
Operations Logs and Reports
24
Procedures
2.0.3
Conduct of Operations
108
Procedures
2.1.10
Station Power Changes
27
Procedures
2.2.68.1
Reactor Recirculation System Operations
Procedures
5.1QUAKE
Procedures
5.3AC120
Loss of 120VAC
Procedures
5.4FIRE-S/D
Fire Inducted Shutdown from Outside Control Room
Work Orders
Corrective Action
Documents
CR-CNS-
25-01839
Miscellaneous
Protected Equipment Tagout CSA-1, CSA and RCIC, Week
2531
Procedures
0-CNS-WM-110
Significant On-line System Outages
Corrective Action
Documents
CR-CNS-
25-03911, 2025-04226
Procedures
0.5 Ops
Operations Review of Condition Reports/Operability
Determinations
Procedures
2.0.5
Reports To NRC Operations Center
Procedures
2.1.11.2
Reactor Building Data
Procedures
2.2.74
Standby Liquid Control System
Procedures
6.LOG.601
Daily Surveillance Log - Modes 1, 2, and 3
2
Drawings
Burns & Roe
201, Sheet 1
Cooper Nuclear Station Turbine Generator Building
Basement - Instrumentation Conduit & Tray Plan
Drawings
Burns & Roe
202, Sheet 2
Cooper Nuclear Station Turbine Generator Building
Basement - Instrumentation Conduit & Tray Plan
Drawings
Burns & Roe
203, Sheet 1
Cooper Nuclear Station Turbine Generator Building
Mezzanine - Instrumentation Conduit & Tray Plan
N28
Drawings
Burns & Roe
Cooper Nuclear Station Control Building Cable & Control
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
213, Sheet 1
Rooms Instrumentation Conduit Plans
Corrective Action
Documents
CR-CNS-
25-04016, 2025-04017, 2025-04202
Procedures
2.2.67
Reactor Core Isolation Cooling System
Procedures
6.FP.101
Fire Pump Functionality Test
Procedures
6.HPCI.103
Procedures
6.RCIC.102
Work Orders
WO 5492996, 5510942, 5511695, 5513155
2.4CRD
CRD Trouble
5.2AIR
Loss of Instrument Air
RPV Level (Failure to Scram)
5.1INCIDENT
Site Emergency Incident
5.1QUAKE
5.3AC120
Loss of 120VAC
5.4FIRE-S/D
Fire Induced Shudown from Outside Control Room
Procedures
RPV Pressurization/Reactor Power (Failure to Scram)
71151
Procedures
0-CNS-LI-114
Regulatory Performance Indicator Process
Corrective Action
Documents
CR-CNS-
25-01839
Procedures
0-BARRIER
Barrier Control
Procedures
5.1FLOOD
Flood
35