IR 05000458/2025004
| ML26034B190 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 02/11/2026 |
| From: | Douglas Dodson NRC/RGN-IV/DORS/PBC |
| To: | Hansett P Entergy Operations |
| References | |
| IR 2025004 | |
| Download: ML26034B190 (0) | |
Text
February 11, 2026
SUBJECT:
RIVER BEND STATION - INTEGRATED INSPECTION REPORT 05000458/2025004
Dear Phil Hansett:
On December 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at River Bend Station. On January 22, 2026, 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.
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 River Bend 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 River Bend 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. 05000458 License No. NPF-47
Enclosure:
As stated
Inspection Report
Docket Number:
05000458
License Number:
Report Number:
Enterprise Identifier:
I-2025-004-0009
Licensee:
Entergy Operations, Inc.
Facility:
River Bend Station
Location:
St. Francisville, LA
Inspection Dates:
October 1, 2025, to December 31, 2025
Inspectors:
T. Betts, Resident Inspector
K. Murphy, Operations Engineer
E. Powell, 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 River Bend 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 Ensure Correct Torque Values for the Lube Oil Cooler Heat Exchanger Flange Bolts for the Emergency Diesel Generator Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000458/2025004-01 Open/Closed None (NPP)71111.15 The inspectors reviewed a self-revealed finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulations Part 50,
Appendix B, Criterion III, Design Control, for the licensees failure to validate and control a foundational design input embedded in a controlling design document. Specifically, the licensee failed to ensure correct torque values for the lube oil cooler heat exchanger flange bolts for the emergency diesel generator design, rendering the division 2 emergency diesel generator inoperable due to an oil leak following a surveillance run on October 20, 2025.
Failure to Follow Procedures Results in Manual Reactor Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000458/2025004-02 Open/Closed
[H.12] - Avoid Complacency 71153 The inspectors reviewed a self-revealed finding of very low safety significance (Green) and an associated non-cited violation of Technical Specification 5.4.1.a for the licensees failure to implement a procedure required by Regulatory Guide 1.33, Revision 2, Quality Assurance Program Requirements, Appendix A, dated February 1978. Specifically, the licensee failed to follow procedure EN-MA-157, Maintenance Configuration Control, and insulate to the extent possible the metal in the electrical cabinet in the main control room within the radius of the lifted lead to prevent an electrical short during performance of a surveillance test.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000458/2025-005-00 Containment Unit Coolers Inoperable Due to Compensatory Measures 71153 Closed LER 05000458/2025-006-00 Potential Loss of Safety Function of Control Building Chillers Due to a Valve Mispositioned 71153 Closed
LER 05000458/2025-007-00 Division 1 Balance of Plant Isolation Results in Manual Reactor Trip 71153 Closed
PLANT STATUS
River Bend Station began the quarter at 100 percent rated thermal power. On October 9, 2025, operators performed a manual reactor scram as a result of a division I balance of plant isolation.
The unit returned to approximately 95 percent rated thermal power on October 12, 2025. On October 13, 2025, the unit reduced power to approximately 75 percent for a planned rod pattern adjustment. The unit returned to 100 percent rated thermal power on October 14, 2025. On October 15, 2025, the unit reduced power to approximately 85 percent for a rod pattern adjustment. The unit returned to 100 percent rated thermal power on October 16, 2025. On November 12, 2025, the unit commenced a planned outage for recirculation pump repairs, reducing power to zero. Power was restored to approximately 94 percent on November 20, 2025. On November 21, 2025, the unit reduced power to approximately 65 percent for a rod pattern adjustment. The unit returned to 100 percent rated thermal power on November 22, 2025. On November 23, 2025, the unit reduced power to approximately 85 percent for a rod pattern adjustment. The unit returned to 100 percent rated thermal power on November 24, 2025. On December 12, 2025, the unit reduced power to approximately 75 percent for a rod pattern adjustment. The unit returned to 100 percent rated thermal power on December 15, 2025, where it remained for the remainder of the quarter.
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)residual heat removal C on November 5, 2025
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the emergency diesel generator C system on October 29, 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)division 3 emergency diesel generator, fire area DG-5, on October 29, 2025 (2)residual heat removal pump C room, fire area AB-4, on November 5, 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 announced fire drill on December 10, 2025.
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated internal flooding mitigation protections in the division 1 and division 2 switchgear rooms in the Control Building on the 98-foot elevation on November 3, 2025.
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03) (1 Sample)
- (1) The inspector reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam administered from August 12, 2025, to September 19, 2025.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated an operator proficiency training graded scenario on October 7, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (3 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1)main steam line isolation pressure transmitter and trip unit on October 6, 2025 (2)instrument air system on October 22, 2025 (3)480 VAC electrical distribution system on November 10, 2025
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:
(1)reactor protection system A motor generator set on October 16, 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)elevated risk during division 1 chilled water system quarterly valve operability test on October 8, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Rosemount hand switches and motor operated valve environmental qualification qualified life pending engineering change on October 28, 2025 (CR-RBS-02205 and CR-RBS-2025-05664)
(2)division 2 emergency diesel generator lube oil leak, 1 percent water in oil from lab test on October 28, 2025 (CR-RBS-2025-05682)
(3)control building chilled water system chiller HVK-CHL1D saturated refrigerant temperature and pressure high out of specification on November 21, 2025 (CR-RBS-2025-05873)
(4)air-operated valve AOVF041A leak repair on December 9, 2025 (CR-RBS-2025-06281)
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)emergency temporary modification to install a wooden plug into condenser waterbox A to stop leak on October 6, 2025
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated the forced outage activities for replacement of the recirculation pump B seal from November 9, 2025, to November 19, 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) (5 Samples)
(1)work order (WO) 54325017, reactor protection system A motor generator set, on October 7, 2025
- (2) WO 54061034, division 3 emergency diesel generator, on October 15, 2025 (3)division 2 emergency diesel generator after lube oil leak repairs on October 27, 2025 (4)reactor water cleanup pump B seal replacement on November 21, 2025 (5)residual heat removal check valve E12-AOVF041A leakage repair testing on November 26, 2025
Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) STP-309-0203, Division III Diesel Generator Operability Test, Revision 345 on October 15, 2025
- (2) STP-309-0202, Division II Diesel Generator Operability Test, Revision 342 on October 20, 2025
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
- (1) STP-255-6302, Div II PVLCS Quarterly Valve Operability Test, Revision 008 on October 22, 2025
- (2) STP-205-6301, LPCS Pump and Valve Operability Test, Revision 031 on November 4, 2025
Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)
(1)observed reactor coolant system leakage surveillance on December 17, 2025
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
- (1) FLX-P2 (WO 54267043) on October 21, 2025
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (2 Samples)
The inspectors evaluated:
(1)drill and exercise performance on October 14, 2025
- (2) drill and exercise performance on October 28, 2025
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification The inspectors verified licensee performance indicator submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (1 Sample)
- (1) October 1, 2024, through September 30, 2025
BI02: RCS Leak Rate Sample (IP Section 02.11) (1 Sample)
- (1) October 1, 2024, through September 30, 2025
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02) (1 Sample)
- (1) The inspectors reviewed the licensees corrective action program to identify potential trends in document control/records management that might be indicative of a more significant safety issue.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02) (3 Samples)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000458/2025-005-00, Containment Unit Coolers Inoperable Due to Compensatory Measures (ADAMS Accession No. ML25293A447)
The inspection conclusions associated with this LER are documented in Integrated Inspection Report 05000458/2025003 under Inspection Results Section 71111.15 (NCV 05000458/2025003-01). This LER is closed.
- (2) LER 05000458/2025-006-00, Potential Loss of Safety Function of Control Building Chillers Due to a Valve Mispositioned (ML25293A255)
The inspection conclusions associated with this LER are documented in Integrated Inspection Report 05000458/2025003 under Inspection Results Section 71111.24 (NCV 05000458/2025003-02). This LER is closed.
- (3) LER 05000458/2025-007-00, Division 1 Balance of Plant Isolation Results in Manual Reactor Trip (ML25342A438)
The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is closed.
INSPECTION RESULTS
Failure to Ensure Correct Torque Values for the Lube Oil Cooler Heat Exchanger Flange Bolts for the Emergency Diesel Generator Design Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000458/2025004-01 Open/Closed None (NPP)71111.15 The inspectors reviewed a self-revealed finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to validate and control a foundational design input embedded in a controlling design document. Specifically, the licensee failed to ensure correct torque values for the lube oil cooler heat exchanger flange bolts for the emergency diesel generator design, rendering the division 2 emergency diesel generator inoperable due to an oil leak following a surveillance run on October 20, 2025.
Description:
On October 20, 2025, during a monthly emergency diesel generator (EDG)surveillance run, operators observed approximately 45 mL/min of lube oil leakage from the floating tube sheet lantern ring of the division 2 EDG (EGS-EG1B) lube oil cooler (EGO-E1B).
As a result of the oil leak, the EDG was secured and declared inoperable, which led to the licensee entering Technical Specification 3.8.1. Following packing replacement, torquing, and hot torquing of the floating tube sheet bolts in accordance with the approved design configuration and procedures, the EDG was returned to operable status on October 24, 2025, and the technical specification limiting condition for operation was exited.
The licensees Engineering Failure Evaluation documented the immediate cause as loosened bolts at the floating tube sheet end as a result of over-compression of the packing, which in turn was caused by initially overtightened bolts that led to extruding the packing and compromising the lantern rings function. The licensee also noted a long-standing use of a non-standard compression scheme with spacers and shims in a configuration unfamiliar to the vendor, a historical practice of applying torque up to approximately 100 ft-lb on the bolts, and recurring leakage history of the lube oil cooler dating back to 1994.
The engineering change (EC) history includes a couple of notable modifications to the EDGs.
Specifically, engineering request (ER)-RB-2000-0284-000 authorized removing a shim from the division 2 EDG to increase packing compression and provided instructions on how to torque the joint per the American Society of Mechanical Engineers (ASME)Section XI work package and control drawing [MSCD-]EGF-EGO*E1B-A, dated April 5, 1985. EC-72841, as well as child ECs EC-72842 and EC-72843 that implemented EC-72841's changes in 2019 for division 1 and division 2 EDG lube oil coolers, respectively, later resized the lantern rings to correct the floating-tube-sheet-to-lantern-ring mismatch and linked the change to prior ERs.
However, the licensee failed to identify bolt torque as a critical characteristic in these changes and therefore failed to re-evaluate torque requirements for the floating-head assembly despite altering compression geometry.
While the inspectors noted that 2019 was a clear opportunity for the licensee to re-evaluate torque requirements for the configuration from 2019 and onward, the inspectors did not identify a clear technical basis for the historical practice of applying torque up to approximately 100 ft-lb, which goes back effectively to the beginning of plant operation.
Vendor correspondence from October 21, 2025, through November 4, 2025, clarified typical practice for packing joints for the lube oil cooler, which included initially torquing the flange bolts to approximately 15 ft-lb with incremental increases up to 50 ft-lb as necessary to stop any leakage. The vendor assessed that the licensee's torque requirements of 100 ft-lb likely led to metal-to-metal flange contact and potentially caused packing damage.
Corrective Actions: The licensee replaced the packing and reassembled the lube oil cooler, which stopped the oil leak. The licensees long-term corrective actions include instituting EC-54339306 to change the torque value on the flange bolts for both division 1 and 2 EDGs based on vendor guidance and performing a performance-needs analysis to address potential training gaps.
Corrective Action References: CR-RBS-2025-05654, CR-RBS-2025-05682, and CR-RBS-2025-05900
Performance Assessment:
Performance Deficiency: The failure to ensure correct torque values for the lube oil cooler heat exchanger flange bolts for the EDG design in accordance with 10 CFR 50, Appendix B, Criterion III is a performance deficiency. Specifically, the licensee failed to ensure correct torque values for the lube oil cooler heat exchanger flange bolts in the emergency diesel generator design documentation, which resulted in the division 2 emergency diesel generator becoming inoperable due to an oil leak following a surveillance run on October 20, 2025.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure correct torque values for the lube oil cooler heat exchanger flange bolts for the EDG rendered the division 2 EDG inoperable due to an oil leak.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2, Section A, Mitigating System Screening Questions, the inspectors determined that the finding screened to a very low safety significance (Green), because
- (1) the finding was a deficiency affecting the design or qualification of a mitigating SSC, but the SSC did not maintain its operability;
- (2) the degraded condition did not represent a loss of the probabilistic risk assessment (PRA) function of a single train TS system for greater than its TS allowed outage time;
- (3) the degraded condition did not represent a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time;
- (4) the degraded condition did not represent a loss of the PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;
- (5) the degraded condition did not represent a loss of a PRA system and/or function as defined in the PRIB or the licensees PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and
- (6) the degraded condition did not represent a loss of the PRA function of one or more non-TS trains of equipment designated as risk-significant in accordance with the licensees maintenance rule program for greater than three days.
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 III, Design Control, requires, in part, that measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the SSCs.
Contrary to the above, from June 6, 2019, until December 15, 2025, measures were not established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the SSCs.
Specifically, the licensee failed to ensure correct torque values for the lube oil cooler heat exchanger flange bolts for the EDG design, which resulted in the division 2 EDG becoming inoperable due to an oil leak following a surveillance run in October 2025.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Follow Procedures Results in Manual Reactor Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000458/2025004-02 Open/Closed
[H.12] - Avoid Complacency 71153 The inspectors reviewed a self-revealed finding of very low safety significance (Green) and an associated non-cited violation of Technical Specification 5.4.1.a for the licensees failure to implement a procedure required by Regulatory Guide 1.33, Revision 2, Quality Assurance Program Requirements, Appendix A, dated February 1978. Specifically, the licensee failed to follow procedure EN-MA-157, Maintenance Configuration Control, and insulate to the extent possible the metal in the electrical cabinet in the main control room within the radius of the lifted lead to prevent an electrical short during performance of a surveillance test.
Description:
On October 9, 2025, Instrumentation and Control (I&C) technicians performed procedure STP-058-4201, "Containment and Drywell Manual Isolation Actuation Logic System Functional Test," and used EN-MA-157, Maintenance Configuration Control, for lifting and landing leads. A technician removed tape from a lifted live lead and inadvertently shorted the lead to the metal enclosure in the cabinet in the main control room, which resulted in a division 1 balance of plant (BOP) containment isolation due to a blown fuse. The BOP containment isolation resulted in the loss of instrument air to the main steam isolation valves (MSIVs) and component cooling water to the reactor recirculation pumps. The BOP containment isolation was unable to be reset by the control room operators, which required a manual reactor scram to be initiated in anticipation of the inboard MSIV's closing on a loss of air and securing the reactor recirculation pumps to prevent potential damage from the loss of cooling.
Corrective Actions: The licensee entered this issue into their corrective action program.
Corrective Action References: CR-RBS-2025-05494
Performance Assessment:
Performance Deficiency: Station procedure EN-MA-157, Maintenance Configuration Control, Section 5.4, Lifted Lead / Cables, Jumpers and Connectors, requires that when lifting/landing leads, then insulate to extent possible within the radius of the lifted lead to prevent electrical shorts. The inspectors determined that the licensees failure to perform the work activity in accordance with EN-MA-157 was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human 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, failing to follow the maintenance configuration control procedure led to an electrical short and blown fuse resulting in a division 1 BOP isolation and a manual reactor scram.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 1, Initiating Events Screening Questions, Section C, Support System Initiators, the inspectors determined the finding required a regional senior reactor analyst to perform a detailed risk evaluation because the finding resulted in an actual loss of a support system (i.e., instrument air and component cooling water) and the insertion of a manual scram. In performing the detailed risk evaluation, the senior reactor analyst determined that the finding associated with the loss of instrument air and component cooling water and reactor scram was of very low safety significance (Green).
The analyst performed an initiating event analysis as called for in Section 8.0, Initiating Event Analyses, of Volume 1, Internal Events, of the Risk Assessment of Operational Events Handbook. The analyst chose to run this analysis as a general plant transient. No other failures beyond the loss of instrument air and component cooling water were assumed to occur. The analyst used the River Bend Station SPAR model, Revision 8.83, on SAPHIRE, Version 8.2.12, with a cutset truncation of 1.0E-12 to calculate the incremental conditional core damage probability attributable to this performance deficiency. The analyst determined that the use of Diverse and Flexible Coping (FLEX) Strategies for Station Blackout (SBO)events should be credited. To credit the use of FLEX, the analyst adjusted the basic event FLX-XHE-XE-ELAP (Operators Fail to Declare Extended Loss of All AC Power (ELAP) When Beneficial) probability to 1.0E-2 for both the nominal and conditional risk cases. The analyst quantified the SPAR model to determine that the baseline conditional core damage probability (CCDP) associated with a general transient event (TRANS) was 1.35E-7.
The analyst modeled the risk impact of the condition attributable to the performance deficiency by adjusting the basic event IE-TRANS (General Plant Transient) to a value of 1.0.
The analyst also assumed that the basic events IAS-MDC-CF-RUN (Plant Compressors Fail from Common Cause to Run), IAS-EDC-FR-1 (Instrument Air Engine Driven Compressor Fails to Run), and CCW-MDP-CF-FR (CCW Pumps Trains Fail from Common Cause to Run)would serve as applicable surrogates for the loss of instrument air and component cooling water due to the containment isolation. Using these assumptions, the analyst performed an initiating event assessment and calculated an estimated CCDP of 4.71E-7. Subtracting the corresponding baseline CCDP referenced above, the analyst determined that the applicable increase in CCDP attributable to this finding was approximately 3.36E-7, which corresponds to a finding of very low safety significance (Green). The analyst assumed that external events would be an insignificant contributor to the increase in core damage frequency because the probability of any external event coinciding with the containment isolation event would be extremely low. The analyst also considered the impact of the finding on large early release frequency (LERF) and determined that risk attributable to LERF was not a dominant metric in the significance determination for this finding. Based on the above considerations, the analyst concluded that the significance of this finding is very low safety significance (Green).
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, I&C technicians failed to implement appropriate error reduction tools to insulate the electrical cabinet within the radius of the lifted lead to account for the recoil of the wire when the tape was removed from the end of the lead.
Enforcement:
Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in NRC Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Section 9.a 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.
The licensee established procedure EN-MA-157, Maintenance Configuration Control, a quality related procedure, to meet the Regulatory Guide 1.33 requirement. Section 5.4 of EN-MA-157 requires the licensee to insulate to the extent possible within the radius of the lifted lead to prevent electrical shorts when lifting/landing leads.
Contrary to the above, on October 9, 2025, the licensee failed to insulate to the extent possible within the radius of the lifted lead to prevent electrical shorts when lifting/landing leads. Specifically, the licensee failed to insulate all the metal in the electrical cabinet in the main control room within the radius of the lifted lead to prevent an electrical short during performance of STP-058-4201, Containment and Drywell Manual Isolation Actuation Logic System Functional Test. As a result, a division 1 BOP containment isolation and manual reactor scram occurred.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Late or Missing Work Orders 71152 The inspectors identified a potential adverse trend with the station's record management and document control. Specifically, the inspectors noted there are numerous condition reports for late or missing work orders across multiple disciplines. For example, the following condition reports document late or missing paperwork:
The resident staff will continue to observe and document any changes to this issue. The licensee documented the inspectors observations in CR-RBS-2026-00282. No findings or violations were identified.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified that no proprietary information was retained or documented in this report.
- On January 22, 2026, the inspectors presented the integrated inspection results to Phil Hansett, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Drawings
PID-08-09A
Engineering P&I Diagram System 309 Diesel Generator
Drawings
PID-08-09D
Engineering P&I Diagram System 309 Diesel Generator
Drawings
PID-27-07C
Engineering P&I Diagram System 204 Residual Heat
Removal LPCI
Procedures
SOP-0031
Residual Heat Removal (SYS #204)
350
Procedures
SOP-0053
System Operating Procedure - Standby Diesel Generator
and Auxiliaries (SYS #309)
347
Procedures
AB-070-504
RHR Pump C Room Fire Area AB-4/Z-1 and Z-2
Procedures
DG-098-052
Diesel Generator Room Fire Area DG-5/Z-1
Procedures
Fire Brigade Drills
Calculations
PN-317
MELC - Max Flood Elevations for Moderate Energy Line
Cracks in CAT 1 Structures
Miscellaneous
Engineering
Report PSA-RBS-
01-IF-FA
River Bend Station Internal Flooding Flood Area Definition
Report
Miscellaneous
Table 03.03-1
River Bend Licensed Operator Annual Operating
Examination Results - 2025
Corrective Action
Documents
CR-RBS-
23-08305, 2023-09008, 2024-01059, 2024-04586, 2025-
01683, 2025-02528, 2025-02536, 2025-02693, 2025-02896,
25-03107, 2025-04748, 2025-05683, 2025-05734
Procedures
Maintenance Rule (A)(1) Process
Procedures
Troubleshooting Control of Maintenance Activities
Work Orders
WO 54077558, 54092097, 54182570, 54255455, 54266840,
266854, 54271798, 54274778, 54279289, 54325017
Procedures
Protected Equipment Postings
Procedures
STP-410-6301
Div I Chilled Water System Quarterly Valve Operability Test
Corrective Action
Documents
CR-RBS-
25-00529, 2025-02205, 2025-05664, 2025-05873, 2025-
05989, 2025-06281
Drawings
PID-27-07A
Engineering P&I Diagram System 204 Residual Heat
Removal LPCI
Engineering
Changes
Leak Seal Injection E12-AOVF041A
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Engineering
Changes
Leak Seal Injection E12-AOVF041A
Miscellaneous
Engineering
Report
CR-RBS-2025-00529 Operability Input
Miscellaneous
Engineering
Report SDC-
2/410
Control Building HVAC System Control Building Chilled
Water System Ventilation Chilled Water System Design
Criteria System Numbers 402 & 410
Miscellaneous
Vendor Manual
216.210-085-
001C
Operating Instructions Manual - Centrifuge Refrigeration
Machine
2
Procedures
Procedure and Work Instruction Use and Adherence
Work Orders
WO 54305389, Task 26
Corrective Action
Documents
CR-RBS-
25-04437
Engineering
Changes
Emergency TMOD: Plug Ejected Anode #17 on Condenser
Waterbox 1A3
Corrective Action
Documents
CR-RBS-
25-05115, 2025-05664, 2025-05989
Drawings
PID-27-07A
Engineering P&I Diagram System 204 Residual Heat
Removal LPCI
Engineering
Changes
Leak Seal Injection E12-AOVF041A
Engineering
Changes
Leak Seal Injection E12-AOVF041A
Miscellaneous
Engineering
Report SEP-RBS-
IST-1
RBS Inservice Testing Bases Document
Procedures
SOP-0090
Reactor Water Cleanup System
Procedures
STP-000-0001
Daily Operating Logs
100
Procedures
STP-204-6501
DIV I ECCS Check Valve Operability Test
Procedures
STP-204-6603
RHR System Refuel Pressure Isolation Valve Test
Work Orders
WO 54305389, Tasks 18 & 20; 54321326, Task 18