IR 05000458/2023003

Integrated Inspection Report 05000458/2023003

ML23310A032
Person / Time
Site:	River Bend
Issue date:	11/13/2023
From:	William Schaup NRC/RGN-IV/DORS/PBC
To:	Hansett P Entergy Operations
Schaup W
References
EA-23-065 IR 2023003
Preceding documents:	ML23349A077 (18 December 2023)
Download: ML23310A032 (44)

Text

November 13, 2023

SUBJECT:

RIVER BEND STATION - INTEGRATED INSPECTION REPORT 05000458/2023003 AND NOTICE OF VIOLATION

Dear Phil Hansett:

On September 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at River Bend Station. On November 13, 2023, 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.

The enclosed report documents a cited Severity Level IV violation for the failure to obtain a license amendment prior to making a change to the licensing basis for the ultimate heat sink for the River Bend Station. The NRC evaluated this violation in accordance with section 2.3.2 of the NRC Enforcement Policy, which can be found at http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html. The violation is cited in Enclosure 1, Notice of Violation (Notice). We determined that this violation did not meet the criteria to be treated as a non-cited violation (NCV), consistent with section 2.3.2 of the Enforcement Policy, because the licensee failed to restore compliance within a reasonable period of time after the violation was identified.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice of Violation when preparing your response. If you have additional information that you believe the NRC should consider, you may provide it in your response. The NRCs review of your response will also determine whether further enforcement action is necessary to ensure your compliance with regulatory requirements.

Additionally, six findings of very low safety significance (Green) are documented in this report.

All of these findings involved violations of NRC requirements, one of which was determined to be severity level IV. One severity level IV violation without an associated finding is documented in this report. We are treating these violations as 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 enclosures, and your response 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, Signed by Schaup, William on 11/13/23 William T. Schaup, Jr., Acting Chief Reactor Projects Branch C Division of Operating Reactor Safety Docket No. 05000458 License No. NPF-47

Enclosures:

1. Notice of Violation 2. Inspection Report 05000458/2023003

Inspection Report

Docket Number: 05000458 License Number: NPF-47 Report Number: 05000458/2023003 Enterprise Identifier: I-2023-003-0012 Licensee: Entergy Operations, Inc.

Facility: River Bend Station Location: St. Francisville, Louisiana Inspection Dates: July 1, 2023, to September 30, 2023 Inspectors: D. Antonangeli, Resident Inspector B. Baca, Health Physicist E. Powell, Resident Inspector H. Strittmatter, Emergency Preparedness Inspector C. Wynar, Senior Resident Inspector Approved By: William T. Schaup, Jr., Acting Chief Reactor Projects Branch C Division of Operating Reactor Safety Enclosure 2

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 Take Risk Mitigating Actions for Unit Cooler A Work Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green H.5 - Work 71111.13 Systems NCV 05000458/2023003-01 Management Open/Closed The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.65(a)(4) for failing to manage an increase in risk from maintenance activities. Specifically, on June 27, 2023, the licensee failed to manage the increase in risk from planned maintenance on the division 1 containment unit cooler.

Failure to Obtain an Amendment for Technical Specification Bases Change Cornerstone Significance/Severity Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.15 Systems Severity Level IV NCV 05000458/2023003-02 Open/Closed The inspectors identified a Green finding and associated severity level IV non-cited violation of 10 CFR Part 50.59, Changes, Tests, and Experiments, when the licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing change package 2000-11 on September 5, 2000. This change added a note to the technical specification bases that changed the intent of the associated surveillance requirement.

Failure to Perform a Radiation Survey in Accordance with 10 CFR 20.1501(a) to Ensure Occupational Doses were Controlled Within Regulatory Limits.

Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.14] - 71124.05 Radiation Safety NCV 05000458/2023003-03 Conservative Open/Closed Bias The inspectors identified a Green non-cited violation for the licensees failure to perform a radiation survey (10 CFR 20.1501(a)) to ensure occupational doses were controlled within regulatory limits (10 CFR 20.1201). Specifically, between July 10 and July 14, 2023, the inspectors identified two examples of area radiation monitors in alert and alarm conditions with dose rate readings that were not reflective of the current area radiological conditions.

Each monitor had reportedly been in this condition at or greater than 29 days. These monitors were installed to provide constant dose rate information locally and/or in the control room and provided local audible and visual alarms upon reaching a preset dose rate.

2-2

Notice of Violation of 10 CFR 50.59, Changes, tests and experiments, for the failure to obtain a license amendment for a change to licensing basis for the River Bend ultimate heat sink.

Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71152A NOV 05000458/2023003-04 Open The inspectors identified a Severity Level IV Notice of Violation of 10 CFR 50.59, Changes, tests and experiments, for the licensees failure to obtain a license amendment for a change to the licensing basis for the ultimate heat sink that required prior NRC approval. Specifically, the licensee failed to seek a license amendment for a change to the facility that allowed the use of makeup to meet the 30-day ultimate heat sink inventory requirements.

Failure to Maintain Accurate Information in Updated Safety Analysis Report Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71152A NCV 05000458/2023003-05 Open/Closed EA-23-065 The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) when the licensee failed to update the updated safety analysis report to assure that the information included in the report contains the latest information developed. Specifically, the licensee failed to ensure that the design capacity of the diesel generator air start system was updated to reflect a change to the diesel engine air start capacity and start time requirements.

Failure to Follow Procedures for Time Critical Actions Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green H.14 - 71152S Systems NCV 05000458/2023003-06 Conservative Open/Closed Bias The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete required actions in accordance with Procedure EN-OP-123,

Time Critical Action Program Standard, revision 007. Specifically, the licensee failed to scope in actions associated with Procedure AOP-0004, Loss of Offsite Power, as time critical and failed to take all time critical actions for shutting SSW-MOV96, Normal Service Water Isolation Valve.

Failure to Properly Categorize Standby Service Water Valves in the Inservice Testing Program Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71153 Systems NCV 05000458/2023003-07 Open/Closed 2-3

The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.55a(f)(4), In-service Testing Requirements, for the licensees failure to categorize standby service water isolation valves as Category A valves. Specifically, the licensees inservice testing program did not test safety-related valves (SWP-MOV-57A and SWP-MOV-57B) in accordance with ASME OM code Subsection ISTC-1300, Valve Categories, to ensure they could meet seat leakage requirements. This caused the licensee to be in violation of Technical Specification 3.7.1 for the operability of the ultimate heat sink.

Engineering Changes Failed to Evaluate Effects on Technical Specification Surveillance Requirements Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.1] - 71153 Systems NCV 05000458/2023003-08 Identification Open/Closed The inspectors identified a non-cited violation of 10 CFR Part 50, appendix B, criterion III and associated technical specification violation of Technical Specification 3.8.1, AC Sources Operating, when the licensee failed to incorporate original circuitry design into a new digital control unit upgrade for the control building air conditioning system. This was identified when the licensee performed STP-309-0602, Division II ECCS Test, revision 056, and the chillers failed to meet the acceptance criteria and load onto the emergency diesel generator in the required time in the final safety analysis report.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000458/2023-001-00 Ultimate Heat Sink 71153 Closed Inoperable due to Boundary Valve Leakage LER 05000458/2023-001-01 Ultimate Heat Sink 71153 Closed Inoperable due to Boundary Valve Leakage (Supplement)

LER 05000458/2023-002-00 Division I and II Diesel 71153 Closed Generators Inoperable due to Exceeding Load Sequence Times LER 05000458/2023-002-01 Division I and II Diesel 71153 Closed Generators Inoperable due to Exceeding Load Sequence Times (Supplement)2-4

PLANT STATUS

The plant began the quarter at 100 percent rated thermal power. On July 11, 2023, the unit was down powered to 85 percent for a planned rod pattern adjustment and returned to 100 percent on July 12, 2023. On August 18, 2023, the unit was down powered to 65 percent for a planned rod pattern adjustment and returned to 100 percent power on August 20, 2023, where it remained for the remainder of the third 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.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (4 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) standby liquid control system on August 30, 2023

(2) high pressure core spray system on August 30, 2023

(3) residual heat removal (RHR) system on September 19, 2023

(4) 125 Vdc system on September 19, 2023

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

(1) cable chase III, fire area C-9, on August 7, 2023

(2) high pressure core spray battery room and battery charger room, fire area C-21, on August 10, 2023

(3) division 1 emergency diesel generator (EDG), fire area DG-6/Z-1, on August 16, 2023 2-5

(4) RHR pump C room, fire area AB-4/Z-1 and Z-2, on September 7, 2023

(5) reactor core isolation cooling (RCIC) pump room, fire areas AB-1/Z-1 and AB-15/Z-1, on September 7, 2023

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

(1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on September 21, 2023.

71111.06 - Flood Protection Measures

Flooding Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated internal flooding mitigation protections in the RHR A cubicle.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)

(2 Samples)

(1) The inspectors observed and evaluated licensed operator performance in the control room during downpower to 85 percent to repair feed pump B oil leak on July 11, 2023.

(2) The inspectors observed and evaluated licensed operator performance in the control room during downpower to 60 percent for rod pattern adjustment on August 21, 2023.

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

(1) The inspectors observed and evaluated an assistant operations manager drill and exercise performance scenario for shift manager proficiency on August 29, 2023.

(2) The inspectors observed and evaluated a graded crew exam on September 19, 2023.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (2 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) division 1 EDG turbo charger bolt failure on August 14, 2023

(2) maintenance preventable functional failures for fire dampers on September 18, 2023 2-6

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (7 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1) elevated risk during RCIC maintenance on July 11, 2023

(2) elevated risk from unplanned limiting condition for operation (LCO) entry for control room fresh air due to division 2 inoperability on July 20, 2023

(3) main turbine rotor crane lift high integrated risk and trip risk on August 2, 2023

(4) elevated risk due to RHR A STP-203-1300, revision 24, new revision to probabilistic risk assessment on August 3, 2023

(5) elevated risk during work on containment A unit cooler on August 7, 2023

(6) elevated risk during RHR A breaker current injection testing on August 7, 2023

(7) elevated risk for containment unit cooler B during standby service water (SSW)

STP-256-6604 on August 22, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)

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

(1) time critical operator actions for SSW-MOV-96B effect on system operability on August 10, 2023

(2) emergency core cooling system drift calculations effect on system operability on August 10, 2023 (CR-RBS-2023-01183, CR-RBS-2023-01184, and CR-RBS-2023-01287)

(3) fire suppression system piping rupture effect on fire suppression operability on August 10, 2023

(4) operability of RHR C shutdown cooling suction line pipe support bend on August 11, 2023 (CR-RBS-2023-05492)

(5) division 2 EDG stator temperatures on August 22, 2023 (CR-RBS-2023-06501)

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)

(2 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) Engineering Change (EC) 93741, upgrade EDG turbo mounting bolts, on August 16, 2023

(2) EC 94390, main steam positive leakage control system MOVF028 replacement, on September 18, 2023

71111.24 - Testing and Maintenance of Equipment Important to Risk

2-7

The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:

Post-Maintenance Testing (PMT) (IP Section 03.01) (7 Samples)

(1) work order (WO) 54013857, RCIC limit switch replacement, on August 11, 2023

(2) WO 00523755, E12-VF063B valve repair, on August 11, 2023

(3) WO 54038407, SWP-MOV507B failed to open, on August 24, 2023

(4) WO 53018691, division 2 control building chilled water system HVC-FS62B, 1HVC-ACU1B flow switch repair, on August 24, 2023

(5) WO 00501472/WO 00574687/WO 00584865, component cooling water secondary pump CCS-P1A repair, on September 8, 2023

(6) WO 52973338, standby liquid control pump 1A discharge header pressure relief valve replacement, on September 8, 2023

(7) WO 54047693, reactor water cleanup pump B tubing replacement, on September 14, 2023

Surveillance Testing (IP Section 03.01) (4 Samples)

(1) STP-403-1301, HVR-UC1B System B Timer Channel Calibration, revision 09, on July 1, 2023

(2) STP-207-4241, RCIC Equipment Room Ambient Temperature High Channel Calibration and LSFT, revision 15, on July 7, 2023

(3) STP-207-4538, RCIC Isolation - RCIC Steam Supply Pressure - Low Channel Functional Test, revision 305, on September 6, 2023

(4) STP-309-0202, DIV 2 EDG Operability Test, revision 339, on September 11, 2023

Inservice Testing (IST) (IP Section 03.01) (2 Samples)

(1) STP-208-6302, Div II MSIV Leakage Control Quarterly Valve Operability Test, revision 12, on September 7, 2023

(2) STP-201-6310, SLC Pump and Valve Operability Test, revision 316, on September 12, 2023

71114.02 - Alert and Notification System Testing

Inspection Review (IP Section 02.01-02.04) (1 Sample)

(1) The inspectors evaluated the licensees maintenance, testing, and audit program for the alert and notification system between July 1, 2021, and August 4, 2023.

71114.03 - Emergency Response Organization Staffing and Augmentation System

Inspection Review (IP Section 02.01-02.02) (1 Sample)

(1) The inspectors evaluated the readiness of the Emergency Preparedness Organization between July 1, 2021, and August 4, 2023. Inspectors also evaluated the licensees ability to staff their emergency response facilities in accordance with emergency plan commitments.

71114.04 - Emergency Action Level and Emergency Plan Changes

2-8

Inspection Review (IP Section 02.01-02.03)

(1) The inspectors evaluated the 10 CFR 50.54(q) emergency plan change process and practices between July 1, 2021, and August 4, 2023. The evaluation reviewed screenings and evaluations documenting implementation of the process. The reviews of the change process documentation do not constitute NRC approval.

71114.05 - Maintenance of Emergency Preparedness

Inspection Review (IP Section 02.01 - 02.11) (1 Sample)

(1) The inspectors evaluated the maintenance of the emergency preparedness program between July 1, 2021, and August 4, 2023. The evaluation reviewed evidence of completing various emergency plan commitments, the conduct of drills and exercises, licensee audits and assessments, and the maintenance of equipment important to emergency preparedness.

71114.06 - Drill Evaluation

Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01)

(1 Sample)

(1) full site emergency preparedness drill with external agency participation on

August 10, 2023 Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)

The inspectors evaluated:

(1) graded crew exam with drill and exercise performance opportunity on August 29,

71124.05 - Radiation Monitoring Instrumentation

Walkdowns and Observations (IP Section 03.01) (8 Samples)

The inspectors evaluated the following radiation detection instrumentation during plant walkdowns:

(1) air samplers (AMS4 and low volume air sampler) in the fuel handling and turbine buildings

(2) area radiation monitors (ARMs) in auxiliary, fuel handling, radwaste, reactor, and turbine buildings

(3) radiation detection and measurement equipment (friskers, high purity germanium detectors, liquid scintillator, proportional counter, portable air sampler) staged for use in the chemistry counting laboratory

(4) portable radiation instrumentation (friskers, iSolo alpha beta counter) staged for use at radiologically controlled area egress 2-9

(5) portable radiation instrumentation (friskers, ion chambers, beta counter, telepoles)staged for use in the instrument storage room

(6) stationary radiation instrumentation (tool and equipment monitors, personnel contamination monitors, and portal monitors) located at the radiologically controlled area egress

(7) stationary radiation instrumentation (portal monitors) at the protected area egress

(8) whole body counter and portal monitor in the dosimetry office

Calibration and Testing Program (IP Section 03.02) (14 Samples)

The inspectors evaluated the calibration and testing of the following radiation detection instruments:

(1) ARGOS-5AB: #140-184 dated April 27, 2023; #140-186 dated October 20, 2022;

1. 1812-369 dated June 23, 2023; and #2206-142 dated January 13, 2023

(2) ASP-1 frisker: CHP-MF-040 dated February 28, 2022, and CHP-MF-124 dated February 28, 2022

(3) CRONOS: #1011-060 dated September 13, 2022; #1011-061 dated November 20, 2022; #1412-364 dated January 25, 2023; and #2205-102 dated February 17, 2023

(4) DMC-3000: #940981 dated February 2, 2023, and A04D33 dated August 17, 2022

(5) Eberline AMS-4: RBS06 dated February 6, 2022, and RBS07 dated September 1, 2022

(6) Eberline BC-4: CHP-BC-001 dated November 8, 2022

(7) Eberline PM7: #419-08 dated August 11, 2022

(8) GEM-5: #1410-188 dated April 21, 2023; #1410-189 dated September 27, 2022;

1. 1902-041 dated November 19, 2022; #2112-320 dated 2023; #2112-321 dated March 23, 2023; and #2206-141 dated February 17, 2023

(9) high purity germanium detectors: Det1 dated March 28, 2023, and Det4 dated March 28, 2023

(10) iSolo: CH-C-012 dated February 6, 2023, and CH-C-066 dated January 1, 2023

(11) Ludlum 144: CHP-CR-075 dated August 4, 2022; CHP-CR-092 dated October 6, 2022; CHP-CR-110 dated August 4, 2022; CHP-CR-118 dated August 4, 2022; CHP-CR-137 dated September 6, 2022; CHP-CR-155 dated October 6, 2022; and CHP-CR-237 dated October 6, 2022

(12) Ludlum 9-3 ion chamber: CHP-DR-487 dated September 13, 2022; CHP-DR-519 dated February 28, 2023; CHP-DR-529 dated February 28, 2023; and CHP-DR-852 dated January 10, 2023

(13) Mirion RDS-31(iTx): CHP-DR-541 dated November 16, 2022; CHP-DR-563 dated December 12, 2022; and CHP-DR-594 dated October 12, 2022

(14) telepole (Tele-Pole and Tele-Pole II): CHP-TEL030 dated November 7, 2022, and CHP-TEL-097 dated August 4, 2022 Effluent Monitoring Calibration and Testing Program Sample (IP Section 03.03) (2 Samples)

The inspectors evaluated the calibration and maintenance of the following radioactive effluent monitoring and measurement instrumentation:

(1) radwaste exhaust (RE-06A/B) and plant stack (RE-126) ventilation systems

(2) liquid effluent (RE-107) monitoring system 2-10

71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &

Transportation

Radioactive Material Storage (IP Section 03.01)

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

(1) storage of radioactive material located within the low level radioactive material building

(2) storage of radioactive material located within the Stone & Webster building

(3) storage of radioactive material located within in the radwaste building

Radioactive Waste System Walkdown (IP Section 03.02) (1 Sample)

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

(1) radwaste resin transfer system

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

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

(1) reactor water cleanup system, waste stream analysis dated 08/09/2022

(2) dry activated waste, waste stream analysis dated 04/17/2023 Shipment Preparation (IP Section 03.04)

There was no sample available when the inspectors were on site.

Shipping Records (IP Section 03.05) (5 Samples)

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

(1) mixed bed ion exchange media, waste class A unstable, LSA-II, shipment number RBS-2022-059

(2) mixed bed ion exchange media, waste class A unstable, LSA-II, Shipment number RBS-2022-034

(3) filter/hardware liner, waste class B, Type B(U), Shipment number RBS-2022-031

(4) mixed bed ion exchange media, waste class A unstable, LSA-II, Shipment number RBS-2023-040

(5) mixed bed ion exchange media, waste class A unstable, LSA-II, Shipment number RBS-2023-041

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification 2-11

The inspectors verified licensee performance indicators submittals listed below:

EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) ===

(1) October 1, 2022, through June 30, 2023 EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13)

(1 Sample)

(1) October 1, 2022, through June 30, 2023 EP03: Alert and Notification System (ANS) Reliability Sample (IP Section 02.14) (1 Sample)

(1) October 1, 2022, through June 30, 2023

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 issue:

(1) heat exchanger inspection on ultimate heat sink (Inspection Report Number 2017-003) on September 20, 2023

71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)

The inspectors reviewed the licensees corrective action program and other licensee trending programs for potential adverse trends that might be indicative of a more significant safety issue. The inspectors provided an observation in the results section with further details of their review.

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 05000458/2023-001-00, Ultimate Heat Sink Inoperable due to Boundary Valve Leakage (ADAMS Accession No. ML23143A276) and LER 05000458/2023-001-01, Ultimate Heat Sink Inoperable due to Boundary Valve Leakage Supplement (ML23229A121)

The inspection conclusions associated with this LER and an associated non-cited violation are documented in this report under IP 71153 in the Inspection Results section. This LER is closed.

(2) LER 05000458/2023-002-00, Division I and II Diesel Generators Inoperable due to Exceeding Load Sequence Times (ML23180A213) and LER 05000458/2023-002-01, 2-12

Division I and II Diesel Generators Inoperable due to Exceeding Load Sequence Times Supplement (ML23256A200)

The inspection conclusions associated with this LER and an associated non-cited violation are documented in this report under IP 71153 in the Inspection Results section. This LER is closed.

INSPECTION RESULTS

Failure to Take Risk Mitigating Actions for Unit Cooler A Work Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green H.5 - Work 71111.13 Systems NCV 05000458/2023003-01 Management Open/Closed The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.65(a)(4) for failing to manage an increase in risk from maintenance activities. Specifically, on June 27, 2023, the licensee failed to manage the increase in risk from planned maintenance on the division 1 containment unit cooler.

Description:

The inspectors identified one example of maintenance activities that were assessed as placing the plant in an elevated risk category but were not appropriately managed in accordance with industry guidance and approved licensee procedures.

On June 27, 2023, the licensee was performing Procedure STP-051-4279, Containment Unit Cooler System Instrumentation, Unit Cooler A - Containment to Annulus Differential Pressure High Channel Calibration and Logic System Functional Test, revision 011. This makes containment unit cooler 1A unavailable and places the plant in a yellow risk condition. During a control room tour, the inspectors noticed that containment unit cooler 1B was not protected.

The inspectors reviewed Procedure EN-OP-119, Protected Equipment Postings, revision 16, and noted that section 5.2.4 states, When the loss of the redundant component or system would result in a technical specification (TS) action statement that requires an immediate plant shutdown (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less) and not performing a surveillance that meets the requirements of step 5.2.9 then protect the component or system.

Procedure EN-OP-119, states, the shift manager is responsible for ensuring the hanging, tracking and prompt removal of protected equipment postings based on the Risk Assessment Process or requirements of Section 5.2, (Conditions that Require Posting of Protected Equipment).

The inspectors determined that the licensee met the conditions in this section of procedure EN-OP-119 that would require them to protect the division 2 containment unit cooler 1B, since loss of the unit cooler 1B would result in a shutdown action statement of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. The licensee failed to take the prescribed risk mitigating actions (RMAs) until questioned by the inspectors.

Corrective Actions: The licensee entered this issue into their corrective action program.

Corrective Action References: CR-RBS-2023-05478 2-13

Performance Assessment:

Performance Deficiency: Title 10 CFR 50.65(a)(4) requires the licensee to assess and manage the increase in risk that may result from maintenance activities before performing them. The inspectors determined that during the maintenance on containment unit cooler 1A system, the licensee failed to assess and manage the increase in risk for maintenance activities by taking required RMAs and was therefore a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to protect the redundant train of containment unit cooler adversely affected the Mitigating Systems cornerstone because the failure to take RMAs for the division 2 containment unit cooler 1B increased the likelihood that other mitigating systems could be affected.

Significance: The inspectors assessed the significance of the finding using IMC 0609 appendix K, Maintenance Risk Assessment and Risk Management SDP. The inspectors requested that the licensee perform a risk assessment of the specific configurations of both conditions. In their assessments, the licensee estimated the risk deficits and incremental core damage probabilities for each condition were less than 1.0E-6. A regional senior reactor analyst independently reviewed the licensees assessments and confirmed the licensees risk estimates. The inspectors applied this information to the flowcharts in appendix K to determine this finding had very low safety significance (Green).

Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the shift manager failed to implement the RMAs required by EN-OP-119.

Enforcement:

Violation: Title 10 CFR 50.65(a)(4) requires, in part, that the licensee to assess and manage the increase in risk that may result from maintenance activities.

Contrary to the above, on June 27, 2023, the licensee failed to assess and manage the increase in risk before performing maintenance activities. Specifically, the licensee failed to implement appropriate RMAs to manage the increase in risk that resulted from maintenance on the containment unit cooler 1A system.

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

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Failure to Obtain an Amendment for Technical Specification Bases Change Cornerstone Significance/Severity Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.15 Systems Severity Level IV NCV 05000458/2023003-02 Open/Closed The inspectors identified a Green finding and associated Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, when the licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing change package 2000-11 on September 5, 2000. This change added a note to the technical specification bases that changed the intent of the associated surveillance requirement.

Description:

The control building air conditioning (AC) system consists of two independent, redundant subsystems that provide cooling and heating of the control building air. The associated control building chilled water subsystem (HVK) supplies chilled water to both subsystem AC units, as well as to the main control room AC units.

The design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the engineered safety feature (ESF) systems. Each ESF bus has a dedicated onsite diesel generator (DG). A DG starts automatically on loss of coolant accident (LOCA) or an ESF bus degraded voltage, undervoltage/loss of power (LOP)signal. Certain required plant loads in divisions 1 and 2 (including HVK) are returned to service in a predetermined sequence to prevent overloading the DG.

The licensee completed modifications to the control building chiller controllers for HVK-CHL 1C and 1D in 2014 and then completed the same modifications to HVK-CHL 1A and 1B in 2020. Historical data recorded shows that since the control building chiller control modifications, the start of the control building chillers may be delayed by approximately 40 seconds past the design start time.

On February 12, 2023, testing of the division 2 emergency core cooling system (ECCS)revealed that both division 2 control building chillers (HVK-CHL 1B and 1D) failed to sequence onto the emergency bus within the TS required time. Investigation of the issue revealed that the same condition was present for the division 1 control building chillers (HVK-CHL 1A and 1C). The licensee determined the condition was caused by the modification that replaced the control building chiller controllers.

The licensee documented this issue in CR-RBS-2023-04966 and performed an operability evaluation for the condition.

In the operability evaluation the licensee referenced a note in the TS bases for surveillance requirement (SR) 3.8.1.18 that states this surveillance requirement pertains only to the load sequence timer itself, and not to the interposing logic which comprises the remainder of the circuit. The evaluation further states that in the case of HVK-CHL 1B, the load sequence timer is relay EJS-SWG1B-62 and is set for a 150 second time delay and that the correct timing of the load sequence timers for the chiller is demonstrated by the start of the air handling units within the 135-165 second band. The note in the TS bases was added by the licensee in change package 2000-11 as part of the 10 CFR 50.59 process after a similar ECCS test failure of the low-pressure core spray pump in 1999.

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During their review of the issue, the inspectors questioned the licensee on the intent of the TS bases note, as well as the 50.59 evaluation performed in 2000 as part of the change package. After receiving the licensees response that the note was added to clarify the SR intent, the inspectors requested assistance from the technical specification branch of the Office of Nuclear Reactor Regulation to provide clarification on the intent of SR 3.8.1.18.

River Bend Station TS SR 3.8.1.18 states Verify sequence time is within +/- 10% of design for each load sequencer timer. The bases states that the intent of SR 3.8.1.18 is that the 10 percent load sequence time tolerance ensures that sufficient time exists for the bus power supply to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated. River Bend Station Updated Safety Analysis Report (USAR) Tables 8.3-2a and 8.3-2b (automatic and manual loading of ESF buses for divisions 1 and 2) include the start time for each ESF load calculated from the time of the LOOP/LOCA signal to start the EDGs.

River Bend Station SR 3.8.1.18 acceptance criteria for the timing sequence intervals should include the start of the associated loads. The function of load sequencing is based on 1) the accident analysis load assumptions and 2) ensuring the EDG has sufficient voltage and frequency to start each load.

The inspectors determined that the added note contradicts the preceding sentence that describes the intent of the surveillance and appears to contradict the licensees current surveillance procedure acceptance criteria. Additionally, the note is also not specific to any load. If it were applied to a load timed earlier than or larger than the control building chiller (which has a relatively long period before the next load would start), it could negatively affect subsequent loads or the EDG itself.

The note is technically incorrect, as it defeats the purpose of meeting the USAR Table 8.3-2b load starting time requirements. Although, this TS bases note did not adversely impact the safety analyses for the chillers because they are the final load sequenced, it can adversely affect the starting of other ESF loads. Any change to SR bases which defeats the real purpose of the SR is incorrect.

Changes to the licensees TS bases are made in accordance with TS 5.5.11, Technical Specifications (TS) Bases Control Program, that provides a means for processing changes to the bases of these TS. The TS states the following:

a. Changes to the bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to bases without prior NRC approval provided the changes do not involve either of the following:

1. A change in the TS incorporated in the license; or 2. A change to the USAR or bases that involves an unreviewed safety question as 2-16

defined in 10 CFR 50.59.

c. The bases control program shall contain provisions to ensure that the bases are maintained consistent with the USAR.

d. Proposed changes that do not meet the criteria of either TS 5.5.11.b.1 or TS 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

The licensee did perform a 50.59 evaluation as part of the change package in July 2000 and as part of that evaluation 10 CFR 50.59(c)(1)(i) required licensees evaluate whether a change to the TS is required prior to making changes to the facility.

The 10 CFR 50.59 evaluation performed by the licensee concluded that no change was required to the TS because the SR 3.8.1.18 timing requirement pertains only to the load sequencer itself and not to the interposing logic which comprises the remainder of the circuit.

This does not compromise the effectiveness of sequencing loads onto the standby DG, therefore, there is no increase in the probability of an accident or a malfunction of equipment important to safety previously evaluated in the River Bend Station safety analysis report.

The inspectors concluded that this change constituted a change to the intent of associated TS and defeated the purpose of meeting the USAR Table 8.3-2b load starting time requirements and can adversely affect the starting of other ESF loads. The added note changes the intent of the testing requirement and defeats the purpose of SR 3.8.1.18, and the licensee incorrectly concluded that the change did not impact the TS and that a license amendment was not required.

Corrective Actions: The preliminary root cause identified that when the controllers were replaced, a delay was introduced in the start logic. This condition resulted in the inability to comply with a SR of TS 3.8.1 and the issuance of licensee event report LER-2023-002. At the time of discovery, River Bend Station was in a mode of operation for which TS 3.8.1 did not apply. Corrective actions to restore TS compliance were completed prior to the next mode of applicability. Additionally, the licensee has a corrective action in place to gather more information to determine a solution that will provide a resolution to address the TS bases note in question.

Corrective Action References: CR-RBS-2023-04060 and CR-RBS-2023-04966

Performance Assessment:

Performance Deficiency: The inspectors determined that the failure of the licensee to obtain a license amendment that impacted the TS prior to making changes to the facility as required by 10 CFR 50.59(c)(2)(ii) was a performance deficiency.

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 licensee failed to obtain a license amendment prior to making a change to the TS bases for SR 3.8.1.18 that changed the intent of the SR and had the potential to adversely affect the starting of ESF loads by the DG.

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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, Mitigating Systems Screening Questions, the inspectors determined that this finding is of very low safety significance (Green) because the impacted SSCs maintained their operability and probabilistic risk assessment (PRA) functionality.

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:

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

Severity: Using the NRC Enforcement Policy, dated January 13, 2023, the inspectors determined the violation was a severity level IV violation in accordance with section 6.1.d.2 because the violation of 10 CFR 50.59 resulted in conditions evaluated as having very low safety significance (Green) by the SDP.

Violation: Title 10 CFR 50.59(c)(2)(ii) requires, in part, that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change that would result in a more than minimal increase in the likelihood of occurrence of a malfunction of a SSC important-to-safety previously evaluated in the USAR.

Contrary to the above, on September 5, 2000, the licensee failed to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change that would result in a more than minimal increase in the likelihood of occurrence of a malfunction of a SSC important-to-safety previously evaluated in the USAR. Specifically, the licensee made a change to the TS bases without obtaining a license amendment that changed the intent of the associated USAR and TS SRs. The licensee failed to recognize that the note added to the TS bases for SR 3.8.1.18, in effect, constituted a change to the intent of associated TS and defeated the purpose of meeting the USAR Table 8.3-2b load starting time requirements.

This type of change would, therefore, require a license amendment pursuant to 10 CFR 50.59(c)(2)(ii).

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

Failure to Perform a Radiation Survey in Accordance with 10 CFR 20.1501(a) to Ensure Occupational Doses were Controlled Within Regulatory Limits.

Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.14] - 71124.05 Radiation Safety NCV 05000458/2023003-03 Conservative Open/Closed Bias The inspectors identified a Green non-cited violation for the licensees failure to perform a radiation survey (10 CFR 20.1501(a)) to ensure occupational doses were controlled within regulatory limits (10 CFR 20.1201). Specifically, between July 10 and July 14, 2023, the inspectors identified two examples of area radiation monitors in alert and alarm conditions with dose rate readings that were not reflective of the current area radiological conditions.

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Each monitor had reportedly been in this condition at or greater than 29 days. These monitors were installed to provide constant dose rate information locally and/or in the control room and provided local audible and visual alarms upon reaching a preset dose rate.

Description:

10 CFR 20.1003, defines a survey as an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.

River Bend Stations USAR, revision 27, section 12.5.2.2.5, established the area radiation monitor (ARM) system in areas where it was desirable to have continuous dose rate information. These monitors indicated dose rate locally and/or in the control room and provided local audible and visual alarms upon reaching a preset dose rate. Therefore, the intent of the ARM system was to provide a continuous survey of radiological conditions of the work areas and inform workers of changing radiological conditions.

During an onsite walkdown of installed plant radiation monitors from July 10-14, 2023, the inspectors identified two ARMs in alert and alarm conditions with dose rate readings not representative of the alert and alarm conditions. The first ARM, RMS-RE-204, Condensate Demineralizer Sink Area, was found to be in an alert condition, i.e., yellow beacon light illuminated. The ARMs displayed dose rate radiation reading was inconsistent with the inspectors independent measurement for activities and materials stored within the local area and was not reflective of the most recent area survey (RBS-2306-00116 dated June 6, 2023).

This area was not an area frequently surveyed by radiation protection (RP) based on the low-level radiological conditions during normal operations. The turbine building, where the ARM was located, was routinely surveyed by RP on a quarterly schedule.

As noted in EIP-2-001, Classification of Emergencies, revision 30, attachment 2 RBS EAL Basis Document, RMS-RE-204 dose rate information was utilized to inform responders (operators, chemistry and radiation protection staff, and fire brigade) of sustained general radiation levels which may preclude access to areas requiring continuous occupancy. In corrective action CR-RBS-2023-05855, dated July 17, 2023, the control room noted a loss of RMS-RE-204 communication and dispatched an operator to review the local conditions. The operator noted the alert beacon was illuminated and radiation dose rate readings were normal. This identification was a week after the inspectors identified RMS-RE-204 was not providing a correct radiation dose rate reading, meaning the operator may not have had an accurate radiation reading on the ARMs radiation dose rate display. The inspectors interviewed an RP technician regarding the alert condition of RMS-RE-204 and was informed the monitor had been intermittently going into alert status for some time and workers were told to disregard the warning indicator. This guidance had the effect of desensitizing workers to radiation alarms.

Based on a review of corrective action documents, the inspectors determined RMS-RE-204 had been having intermittent issues, i.e., reaching an alert/alarm condition without valid conditions present, having erroneous radiological readings, and loosing communication with the control room, since 2011 (CR-RBS-2011-2573). The inspectors reviewed corrective action documents linked to the current alert condition and identified CR-RBS-2021-03740 dated May 19, 2021, as the most recent corrective action document for the monitors issues.

The licensee confirmed no additional corrective actions were generated for RMS-RE-204 since 2021. The inspectors requested confirmation of compensatory continuous dose rate monitoring for the local area. The licensee demonstrated compensatory continuous dose rate monitoring was established on November 15, 2021. This means the area was without continuous dose rate monitoring for approximately 180 days. However, the area was 2-19

surveyed for dose rates during the quarterly surveys dated July 14, 2021 (RBS-2107-00140)and October 27, 2021 (RBS-2110-00229). These surveys provided an assessment of the area 56 days after the ARM was identified deficient and 19 days prior to the compensatory monitoring being established with 105 days in between the quarterly surveys the ARM was deficient. Although quarterly surveys were performed and no abnormal conditions were identified, RMS-RE-204 was not performing its intended function to provide accurate continuous dose rate monitoring and informing workers of elevated radiation dose rates through alert and alarm warning indications.

The compensatory monitoring was provided through the installation of an assigned DMC-3000 teledosimeter (dosimeter number 912967) co-located at RMS-RE-204s detector. The teledosimeter transmitted data via the ATOMS telemetry server to monitoring stations at the radiologically controlled area access point. In addition, the teledosimeter readings could be remotely accessed through the licensees network, as needed. The licensee performed WO 52986825-01 dated December 12, 2021, to correct the monitor issue. However, the intermittent conditions returned in the subsequent months based off an interview with an RP technician and the inspectors observation during another facility walkdown during March 13-16, 2023.

The second ARM observed during the onsite inspection and facility walkdown, RMS-RE-162, 67-foot Elevation Offgas Regeneration Area, was found in an alarm condition (audio alarm engaged with no warning indication lights). The inspectors were unable to validate the dose rate readings of the ARM due a low light environment, i.e., confirm the ARM was functional.

However, the inspectors independent measurement was not consistent with the ARMs warning indication as the dose rates measured by the inspectors were near background levels. The inspectors interviewed an RP technician regarding the length of time this monitor was in the alarm condition. The RP technician believed the monitor had been in this condition for approximately a month. The licensee provided CR-RBS-2023-04658 dated May 28, 2023, for RMS-RE-162s current alarm condition with no indicator (alert or alarm) lights observed.

This means the ARM was in an alarm condition for approximately 42 days when the inspectors found the monitor in alarm. The licensee generated work request WR 514828 dated May 28, 2023, to resolve the issue. The offgas regeneration area was routinely surveyed by RP on a quarterly basis due to the low levels of radiation during normal operations. The most current radiological survey was performed on June 26, 2023 (RBS-2306-00400) and radiation dose rates were as anticipated, i.e., close to background levels.

However, this condition provided 29 days the area was not continuously monitored for changes to radiation dose rates nor provided a valid local audible and visual alarm upon reaching a preset dose rate, i.e., RMS-RE-162 was not performing its intended function.

Further, the continued audio alarm desensitized workers to valid warning indications.

The inspectors reviewed applicable corrective action documents associated with RMS-RE-162 and determined this monitor was experiencing false elevated dose rate readings which triggered alerts and alarms since 2005 (CR-RBS-2005-1546). The latest troubleshooting and resolution to the intermittent issues with RMS-RE-162 was performed under WO 52989527-01 dated July 17, 2022. The intermittent alarms returned as the radiation monitor was found in alarm by the inspectors.

Corrective Actions: The licensee initiated the troubleshooting WOs to address the intermittent alert and alarm conditions of RMS-RE-162 and RMS-RE-204, as well as the communication losses to the control room. Compensatory monitoring was only initiated for RMS-RE-204 after six months.

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Corrective Action References: The conditions were entered into the corrective action program as CR-RBS-2021-03740, CR-RBS-2023-04658, and CR-RBS-2023-05855.

Performance Assessment:

Performance Deficiency: The failure to perform a radiation survey in accordance with 10 CFR 20.1501(a) to ensure occupational doses were controlled within regulatory limits was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to perform radiological surveys increased the workers potential exposure to unknown and/or elevated radiation levels in the work area and desensitized the workers to radiation alarms.

Significance: The inspectors assessed the significance of the finding using IMC 0609 appendix C, Occupational Radiation Safety SDP. The inspectors determined the finding to be of very low safety significance (Green) because

(1) it was not associated with as low as is reasonably achievable (ALARA) planning or work controls,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee relied on operator and RP knowledge and recent quarterly surveys of the affected areas to understand the current and present radiological conditions of these areas until corrective actions to resolve the alerts and alarms were completed. This led workers to rely on their electronic alarming dosimetry for unsuspected radiological changes when in these areas rather than the continuous dose rate monitoring and warning indications of the ARMs.

Enforcement:

Violation: Title 10 CFR 20.1003 defines a Survey as an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.

Title 10 CFR 20.1501(a) requires in part, that each licensee make or cause to be made surveys of areas that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20, and that surveys are reasonable under the circumstances to evaluate the magnitude and extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present.

Title 10 CFR 20.1201(a) requires, in part, that the licensee shall control occupational doses to individual adults less than the annual limit of 5 rem total effective dose equivalent.

River Bend Stations USAR, revision 27, section 12.5.2.2.5, established the ARM system in areas where it was desirable to have continuous dose rate information. These monitors indicated dose rate locally and/or in the control room and provided local audible and visual 2-21

alarms upon reaching a preset dose rate. Therefore, the intent of the ARM system was to provide a continuous survey of radiological conditions of the work areas and inform workers of changing radiological conditions.

Contrary to the above, from approximately May 19 through November 15, 2021, and May 28 through June 26, 2023, the licensee failed to make or cause to be made, with two examples, surveys of areas that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20; and are reasonable under the circumstances to evaluate the magnitude and extent of radiation levels. Specifically, the licensee did not provide compensatory continuous dose rate surveys when the ARM system, in areas where it was desirable to have continuous dose rate information, was unable to provide indications of dose rates locally and/or in the control room and provide local audible and visual alarms upon reaching a preset dose rate.

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

Notice of Violation of 10 CFR 50.59, Changes, tests and experiments, for the failure to obtain a license amendment for a change to licensing basis for the River Bend ultimate heat sink.

Cornerstone Severity Level Cross-Cutting Report Aspect Section Mitigating Severity Level IV Not Applicable 71152A Systems NOV 05000458/2023003-04 Open The inspectors identified a Severity Level IV Notice of Violation of 10 CFR 50.59, Changes, tests and experiments, for the licensees failure to obtain a license amendment for a change to the licensing basis for the River Bend ultimate heat sink that required prior NRC approval.

Specifically, the licensee failed to seek a license amendment for a change to the facility that allowed the use of makeup to meet the 30-day ultimate heat sink inventory requirements.

Description:

The inspectors reviewed corrective actions associated with non-cited violation (NCV)05000458/2011008-06 of 10 CFR 50.59, Changes, Tests, and Experiments, and identified that the licensee did not restore compliance with the requirements of 10 CFR 50.59.

Specifically, the licensee failed to correct the NCV and restore compliance when they closed corrective actions for the violation without obtaining a license amendment.

In October 2011, inspectors reviewed River Bend Stations USAR, section 9.2.5, Ultimate Heat Sink, which describes the standby cooling tower and water storage basin that functions as the ultimate heat sink (UHS) for River Bend Station during accident conditions. USAR section 9.2.5.1, Design Bases, describes the criteria to which the UHS is designed in accordance with General Design Criteria 44, Cooling Water. In particular, USAR section 9.2.5.1, criterion 2 states: The capacity of the [Ultimate Heat Sink] water storage basin is designed to provide necessary cooling for the period of time (30 days) needed to evaluate the situation, to take corrective action to mitigate the consequences of an accident, and if required to take any necessary measures to permit water replenishment. In addition, alternate methods are available for ensuring the continued capability of the sink beyond 30 days (section 9.2.5.2).

Criterion 6 states: The UHS is designed to perform its intended safety function assuming any single active or passive failure coincident with a loss of offsite power.

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When River Bend Station was originally licensed in November 1985, the licensee submitted the design of the UHS that supplied a 30-day supply of water from the standby cooling tower basin for decay heat removal, without a makeup water source. The NRC reviewed and approved this design using Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants. The NRCs position was reflected in the River Bend Station Safety Evaluation Report, NUREG-0989, section 9.2.5, which states, the UHS contains more than a 30-day supply of water for decay heat removal without makeup, in accordance with General Design Criteria 44.

Makeup water required after 30 days of UHS operation can be provided from the non-safety related makeup water system if this system is available.

The original submitted Final Safety Analysis Report, revision 0 had the same criterion for 2 and 6 as it is presently stated in the current USAR revision.

Additionally, the UHS design of a standby cooling tower basin containing a 30-day supply without makeup is confirmed in the River Bend Station Technical Specification Bases, section B 3.7.1, Standby Service Water System and Ultimate Heat Sink. Page B 3.7-1 states, the basin is sized such that sufficient water inventory is available to provide heat removal capability to safely shut down the plant and to maintain it in a cold shutdown condition for a 30-day period with no external makeup water source available (Regulatory Guide 1.27, Ref.

1).

The licensee intended to correct NCV 05000458/2011008-06 by submitting license amendment request RBS LAR-2013-0018 that would credit makeup to the ultimate heat sink in less than 30 days to account for system leakage and operation with more than one division of standby service water in operation; however, the request was later withdrawn on July 7, 2014, due to NRC concerns. This license amendment was intended to correct the 2011 50.59 NCV and to restore compliance.

The following is an excerpt from the NRC response:

The NRC staff has concluded that the licensee did not provide sufficient information to support its position that UHS/SSW leakage is beyond the design basis. In its letter dated May 29, 2014, the licensee cited NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, section 9.2.1, Station Service Water System, as part of the basis for its position that leakage is not a criterion for design of the UHS. However, SRP section 9.2.1 also states that the SWS shall have the capability to isolate components, subsystems, or piping if required so that the system safety function will not be compromised. As such, the NRC staff believes that the licensees position appears contrary to the guidance in SRP section 9.2.1. Therefore, if Entergy decides to re submit the request, it must include the following information:

1) Provide the basis for why accounting for leakage, which compromises a safety function, is not part of RBSs design basis for UHS inventory.

2) If leakage is determined to be part of the licensing basis, specify a leakage limit from the SSW to NSW.

3) Provide the basis for why the operation of one division of ECCS and SSW is the bounding single failure when determining UHS inventory requirements.

4) From the submittals provided, the NRC staff could not verify that the alternative sources 2-23

of makeup water to the UHS provide acceptable methods to replenish the UHS. Provide more detail on the three alternate sources of makeup water with regards to capacity, implementation, and design.

A 2014 component design basis inspection reviewed the licensees corrective actions to address NCV 05000458/2011008-06 and issued a NCV against 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, because the licensee failed to take prompt corrective actions to restore compliance for the 2011 violation.

As part of corrective actions from the 2011 and 2014 NCVs, the licensee revised the Final Safety Analysis Report (FSAR), Section 9.2.5, to remove the following sentence describing the licensing basis of the UHS:

Additional makeup is required for system leakage under licensing basis condition and when operating two divisions with system leakage.

While removal of this sentence addressed a portion of the FSAR changes that caused the 10 CFR 50.59 NCV issued in 2011, the inspectors determined that additional portions of the RBS licensing basis still relied on the use of make-up to meet the 30-day requirement for the UHS and that the licensee had still not received NRC approval for this change. Specifically, FSAR Section 9.2.5.1.2 states that, the capacity of the UHS water storage basin is designed to provide necessary cooling for the period of time (30 days) needed to evaluate the situation, to take corrective action to mitigate the consequences of an accident, and if required to take any necessary measures to permit water replenishment. This is different than the approved section from the SER that stated the UHS had more than a 30-day supply with no makeup required. The statement above has a section to allow makeup if required i.e., cant make the 30-day time as well as a beyond 30-day makeup statement. Further, the inspectors noted that AOP-004, Loss of Offsite Power, still has section 5.13.3 and a note to provide makeup at day 10 of an accident.

The inspectors reviewed the current calculation of record, G13.18.13.2*086, Effects of Maximum Safeguards Operation on the Ultimate Heat Sink (Standby Cooling Tower),revision 3 issued in September 2018 and noted that for all design-basis scenarios, the station can maintain SSW inlet temperature under 95 degrees Fahrenheit provided they take the procedural actions in the USAR and AOP-004, Loss of Offsite Power. These actions are based on the previous 10 CFR 50.59 changes that were the subject of NCV 05000458/2011008-06. The calculation states that, if both divisions of standby cooling tower fans start on initiation of the event and no EDG failure occurs, even if the actions in AOP-004 are taken, the inventory will only last 21.4 days. This duration also assumes no SSW inventory loss through SWP-MOV 96A/B from a delayed closure. When a bounding loss of SSW is assumed through SWP-MOV 96A/B, where one of the normal service water (NSW) to SSW isolation valves fails to close remotely and requires local operation (the worst case single active failure), UHS inventory lasts for 20.4 days.

In summation, the inspectors determined that the licensee failed to restore compliance to NCV 05000458/2011008-06 because they failed to obtain a license amendment for a change to licensing basis for the River Bend ultimate heat sink that required prior NRC approval. In particular, the original RBS license basis stated that the UHS will last 30 days assuming single active failure with no makeup required. However, the current design basis calculation G13.18.13.2*086, revision 3 shows inventory will not last 30 days and the current USAR and AOP-004 allow the licensee to makeup to the UHS to meet the 30-day requirement. The 2-24

allowance for makeup to the UHS was done so without prior NRC approval, contrary to 10 CFR 50.59.

Corrective Actions: The licensee entered this issue into their corrective action program.

Corrective Action References: CR-RBS-2023-07894 and CR-RBS-2023-02086

Performance Assessment:

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

Enforcement:

The Reactor Oversight Process SDP does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter noncompliance.

Severity: Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, it is necessary to address this violation using traditional enforcement to adequately deter noncompliance. Using the NRC Enforcement Policy, dated January 13, 2023, the violation was determined to be Severity Level IV in accordance with section 6.1.d.2 because it involved a violation of 10 CFR 50.59 where the conditions associated with the inappropriate change were evaluated as having very low safety significance (i.e., green) by the SDP.

Violation: Title 10 CFR 50.59(c)(2)(ii) requires, in part, that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change if the change would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important-to-safety previously evaluated in the final safety analysis report (as updated).

Contrary to the above, from October 27, 2011, to September 30, 2023, the licensee failed to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if this activity would result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the final safety analysis report (as updated). Specifically, the licensee changed the design basis of the ultimate heat sink inventory requirements from providing a 30-day cooling water supply without the need for makeup to providing a less than 30-day cooling water supply with makeup capability, without obtaining a license amendment. The licensee implemented temporary compensatory measures that would ensure a 30-day ultimate heat sink cooling water supply.

Enforcement Action: This violation is being cited because the licensee failed to restore compliance within a reasonable period of time after the violation was identified, consistent with section 2.3.2 of the Enforcement Policy.

Failure to Maintain Accurate Information in the Updated Safety Analysis Report Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71152A Applicable NCV 05000458/2023003-05 Applicable Open/Closed EA-23-065 2-25

The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) when the licensee failed to update the updated safety analysis report to assure that the information included in the report contains the latest information developed. Specifically, the licensee failed to ensure that the design capacity of the diesel generator air start system was updated to reflect a change to the diesel engine air start capacity and start time requirements.

Description:

During a review of the DG air start system, the inspectors noted that the design capacity of the starting air banks for the standby DG was described in the USAR as capable of starting the standby DG eight times with five of the starts each within 10 seconds.

Specifically, section 9.5.6 of the USAR revision 27 states:

Each redundant DGSS train is capable of providing the standby diesel generator with eight starts (five of them are 10 sec starts) from two air receivers without recharging the associated air receivers.

While the inspectors were researching the licensing basis for the River Bend Station DGs, the licensee issued EC 93806 to modify USAR section 9.5.6.1.5 to replace the original design basis of the DG air start system with a revised design basis.

The revised design basis (USAR revision 27) stated:

Each redundant DGSS train is capable of providing the standby diesel generator with five consecutive starts from two air receivers without recharging the associated air receivers.

It initially appeared that this change was made without prior NRC approval. During follow-up inspection, the inspectors identified that the NRC approved a change to the diesel air starting system that corresponded to the change that was implemented under EC 93806 with Amendment No. 91 to Facility Operating License No. NPF-47, on January 16, 1997.

However, the licensee did not update the USAR until the time of the inspection in 2022.

Corrective Actions: The licensee entered this issue into their corrective action program.

Corrective Action References: CR-RBS-2023-07903

Performance Assessment:

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

Enforcement:

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

Severity: Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, it is necessary to address this violation using traditional enforcement to adequately deter noncompliance. In accordance with NRC Enforcement Policy, section 6.1.d.3, the violation was determined to be severity level IV because the licensee failed to update the USAR as required by 10 CFR 50.71(e) and the lack of up-to-date information had a material impact on safety or licensed activities.

Violation: Title 10 CFR 50.71(e) requires, in part, licensees shall update periodically, as provided in paragraphs (e)(3) and

(4) of 10 CFR 50.71, the USAR to assure that the information included in the report contains the latest information developed.

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Contrary to the above, from January 16, 1997, to September 30, 2023, the licensee failed to update the USAR to assure that the information included in the report contains the latest information developed. Specifically, the licensee failed to update the FSAR to reflect the NRCs approval of River Bend Station, Amendment No. 91 to Facility Operating License No.

NPF-47, issued on January 16, 1997.

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 for Time Critical Actions Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green H.14 - 71152S Systems NCV 05000458/2023003-06 Conservative Open/Closed Bias The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete required action in accordance with procedure EN-OP-123, Time Critical Action Program Standard, revision 007. Specifically, the licensee failed to scope in actions associated with Procedure AOP-0004, Loss of Offsite Power, as time critical and failed to take all time critical actions for shutting SSW-MOV96, Normal Service Water Isolation Valve.

Description:

While investigating the UHS and SSW systems, the inspectors discovered two specific cases where the licensee had not followed all of the requirements of EN-OP-123. The procedure defines time critical actions (TCAs) as a manual action or series of actions, the performance of which within specified time constraints has been credited in the plant deterministic design analysis or licensing basis. Section 5.1 requires a list of all sources of plant-specific TCAs including the Updated Final Safety Analysis Report (USAR).

The first scenario is related to the SSW to NSW isolation valves failing to shut during an event. Section 9.2.5 of the USAR is for the UHS and 9.2.5.2, system description, describes the importance of shutting either SSW-MOV96 A or B, the NSW to SSW isolation valves, in 20 minutes to mitigate the loss of SSW inventory to the non-safety related NSW system.

6 of EN-OP-123 lists verifying the closure of these valves as a TCA; however, during the review of the associated TCA conducted on October 17, 2020, the inspector the inspector identified that the licensee was not validating the local operation of SSW-MOV96 A.

The licensee, in response to questions from the inspector performed a validation, on May 17, 2023, for the local operation of SSW-MOV96 A. During the validation the inspector identified that multiple different steps of the procedure were not completed for this TCA. Specifically, EN-OP-123, section 5.6, Validation Requirements, step 3, states to ensure any special equipment, devices, or supplies required to support the TCA are readily available to operate SSW-MOV96A.

Specifically, the inspector identified that a step stool or ladder would be required for an operator to operate the valve. Not having the required equipment staged nearby to complete the task results in the 20-minute requirement not being met by the time the proper equipment was retrieved. Additionally, the licensee did not calculate the time required for an operator to arrive on station or operate the valve under actual system conditions as required per sections 2-27

4.3.b.5, 5.6.3, 5.7.3.b, 5.7.5, and 5.8.4. Additionally, section 5.2 step 9 was also not completed which requires a rim pull calculation to be done when stroking the valve under system pressure is not practical.

The second scenario is related to the actions the licensee must take to conserve UHS inventory to complete the 30-day mission time and prevent the UHS water basin from exceeding the design limit of 95 degrees Fahrenheit. Section 9.2.5.3 is the safety evaluation section in the USAR for the UHS. This section required various actions to be completed in 1-2 hours if both divisions of diesels and SSW operate on a LOOP combined with a LOCA. For this case, the USAR states that various actions need to be taken such as throttling flow to the RHR heat exchangers. Additionally, abnormal operating procedure AOP-004, Loss of Offsite Power, revision 73, has section 5.13, Standby Service Water Operations, that has multiple actions to be completed to conserve UHS inventory when both diesels and SSW systems start. Examples include step 5.13.2 requiring operators to secure the division 2 RHR heat exchanger and containment unit cooler as well as throttling flow to the division 1 RHR heat exchanger. None of the actions associated to both safety trains starting at an event were scoped into the TCAs and governed by EN-OP-123.

Corrective Actions: The licensee entered this issue into their corrective action program.

Corrective Action References: CR-RBS-2023-00543, EC 54034873

Performance Assessment:

Performance Deficiency: The licensee failed to scope actions in the USAR and AOP-0004 for preserving UHS water inventory for its 30-day mission time and keeping the basin temperature under 95 degrees Fahrenheit as TCAs. Additionally, the licensee failed to take all of the actions quality procedure EN-OP-123 required for the TCA of closing SWP-MOV96A(B) within 20 minutes of failing to close. Not applying TCA standards to securing division 2 loads and not completing all required steps for TCAs is within the licensees ability to foresee and correct and therefore a performance deficiency.

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.

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, Mitigating Systems, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of the system, did not represent a loss of the PRA function of a single train TS system for greater than its TS-allowed outage time, did not represent a loss of 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 />, did not represent a loss of a PRA system and/or function as defined in the PRIB or licensee's 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 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 licensee's maintenance rule program for more than 3 days.

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is 2-28

determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee failed to identify that the required actions of EN-OP-123 could be performed.

Enforcement:

Violation: Title 10 CFR 50, appendix B, criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be described by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. The licensee established Procedure EN-OP-123, Time Critical Action Program Standard, as the implementing procedure for establishing TCAs that are assumed to be completed within specified time limits and periodically validates and documents the plant staff capability to perform such actions within specified time limits and tracks actual performance times, an activity affecting quality.

Contrary to the above, from September 27, 2022, to September 30, 2023, for TCAs, an activity affecting quality, the licensee failed to describe by a documented procedure, and failed to accomplish the activity affecting quality in accordance with a procedure. Specifically, for procedure EN-OP-123, the TCA implementing procedure, the licensee failed to validate the proper equipment was staged as directed by step 3 of section 5.6, failed to calculate rim pull as directed by step 9 of section 5.2, and failed to scope in TCAs directed by the USAR and AOP-004 into the TCA program as directed by section 5.1.

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

Observation: Improper Storage of Equipment in Safety-Related Buildings 71152S While conducting routine plant status walkdowns, the inspectors made frequent observations of loose tools, equipment, hoses, etc. being left in various locations in safety-related buildings. The licensee utilizes work in progress (WIP) signs to identify temporary services and equipment (TSE) in the plant supporting active work activities. These materials often had expired WIP signs associated with them or incomplete or inaccurate information on the WIP signs. When brought to the attention of the control room, condition reports (CRs) were generated. Occasionally, the equipment would be picked up and stored in its proper location, but the usual fix was to revise the removal dates on the WIP signs to a new date 90 days out.

Ninety days is the maximum amount of time equipment can be left out without further evaluation. By continually updating the WIP with new removal dates, multiple staged equipment locations had material in place that far exceeded the 90-day time without any evaluation. The residents identified at least ten CRs describing locations where there were either no WIPs, TSEs were improperly restrained to safety-related piping, or the TSE there had exceeded 90 days.

The residents determined two specific examples were of note. In the control building AC rooms, there are two safety-related valves, one in each division, SWP-PVY32A/B. These valves have associated CRs describing leakage. The valves have a catch installed underneath them with hoses routed to collection barrels. On multiple occasions the barrels have been identified to either not be restrained or restrained to safety-related piping. The WO numbers referenced on the WIP signs were for different valves. Additionally, the catch, hose, and drum meet the definition of TSE and have been installed in the plant in excess of two years. When the WIP sign expired, the licensee updated the sign with a new 90-day interval.

ADM-0073 requires an evaluation as a temporary modification or permanent change.

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NUMARC 93-01 requires a 50.59 evaluation for alterations related to a maintenance activity that exceeds 90 days.

None of the issues rose to more than minor and the resident staff plans to continue to observe and document any changes to this issue.

Failure to Properly Categorize Standby Service Water Valves in the Inservice Testing Program Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71153 Systems NCV 05000458/2023003-07 Open/Closed The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.55a(f)(4), In-service Testing Requirements, for the licensees failure to categorize standby service water isolation valves as Category A valves. Specifically, the licensees inservice testing program did not test safety-related valves (SWP-MOV-57A and SWP-MOV-57B) in accordance with ASME OM code Subsection ISTC-1300, Valve Categories, to ensure they could meet seat leakage requirements. This caused the licensee to be in violation of Technical Specification 3.7.1 for ultimate heat sink inoperable.

Description:

As documented in NRC inspection report 05000458/2022-001 (ML22112A213),the licensee received a non-cited violation for the failure to correct a condition adverse to quality. Specifically, the NRC reviewed CR-RBS-2019-06296 which documented an instance where the SSW C discharge check valve SWP-V148 did not fully seat after a periodic test of the C SSW pump. The condition was noticed by licensee staff after they observed NSW system leaking by to the SSW system. Check valve SWP-V148 should fully seat in all conditions to prevent leak-by that could divert system flow from SSW. The NRC issued the non-cited violation for the licensees failure to correct the leaking SWP-V148 valve for nearly three years.

The NRC inspectors questioned why other isolation valves within the SSW system were not tested for leakage given that the design basis calculation PM-194, Standby Cooling Tower Performance and Evaporation Losses Without Drywell Coolers, revision 11, identifies that the maximum allowable leakage for the entire SSW system is 6.9 gallons per minute. After researching the issue, the inspectors found that since the year 2000, River Bend Station documented (CR-RBS-2000-01977) the potential for SSW inventory loss through boundary valves that encroached on the 10 gallon per minute limit specified in PM-194, revision 5 (which was the revision of record at the time). Further, to date, the licensee does not have a clearly defined population of SSW boundary valves or what the actual leakage rates are (only the limit of 6.9 gallons per minute).

As stated above, calculation PM-194 assumed that the worst-case leakage from all SSW boundary valves was 6.9 gallons per minute; however, not all the SSW isolation valves have been tested for seat leakage. The licensee incorrectly assumed that inservice testing requirements did not apply to valves SWP-MOV-57A and SWP-MOV-57B, because there were check valves in the flow path. In other words, the license took credit for the redundant check valves, though it is possible the check valves themselves could leak, to exclude valves SWP-MOV-57A and SWP-MOV-57B from the inservice testing program.

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Corrective Actions: In response to this issue, the licensee performed maintenance on SWP-MOV-57A and SWP-MOV-58B to correct the seat leakage. Additionally, the licensee re-categorized valves SWP-MOV-57A and SWP-MOV-58B as Category A valves, updated the inservice testing requirements, and established preventative maintenance tasks for the valves. Finally, the licensee is developing an engineering assessment to ensure that all the SSW boundary valves are correctly categorized in the inservice testing program and tested for leakage. The current worst-case leakage rates of SSW valves do not challenge the available water inventory for over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Corrective Action References: CR-RBS-2023-01249

Performance Assessment:

Performance Deficiency: The failure to correctly test and classify safety-related valves SWP-MOV-57A and SWP-MOV-58B as Category A in accordance with 10 CFR 50.55(a)(f)(4)was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to test for valve leakage on valves SWP-MOV-57A and SWP-MOV-58B adversely affected the reliability of the SSW system.

When tested, SSW valves SWP-MOV-57A and SWP-MOV-58B had leakage in excess of 40 gallons per minute which challenged the available water inventory in the UHS. This resulted in the inability for it to meet its 30-day mission time and TS requirements.

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, Mitigating Systems, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of the system, did not represent a loss of the PRA function of a single train TS system for greater than its TS-allowed outage time, did not represent a loss of 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 />, did not represent a loss of a PRA system and/or function as defined in the PRIB or licensee's 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 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 licensee's maintenance rule program for more than 3 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 50.55a(f), Inservice Testing Requirements, subsection

(4) requires, in part, that pumps and valves which are classified as ASME Class 1, Class 2, and Class 3 must meet the inservice test requirements set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section.

Furthermore, subsection (f)(4)(ii) requires inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of 2-31

the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 18 months before the start of the 120-month interval. ASME Code subsection ISTC-1300, Valve Categories, requires, in part, that valves within this subsection shall be placed in one or more of the following categories. Category A is for valves for which the seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function as specified in ISTA-1100.

TS 3.7.1, Standby Service Water System (SSW) and Ultimate Heat Sink (UHS), applicable in modes 1, 2, & 3, condition D requires that the UHS shall be operable.

Contrary to the above, from November 2000, to September 30, 2023, the licensee failed to categorize SSW valves SWP-MOV-57A and SWP-MOV-58B as Category A valves in accordance with the ASME code requirements. Specifically, by categorizing these valves as B their leakage was not tested and accounted for to ensure UHS operability. After NRC questioning, it was discovered the valves leaked more than 40 gallons per minute. This resulted in the UHS being inoperable for a time longer than permitted by TS, which required the plant to be in mode 3 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />.

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

Engineering Changes Failed to Evaluate Effects on Technical Specification Surveillance Requirements Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.1] - 71153 Systems NCV 05000458/2023003-08 Identification Open/Closed The inspectors identified a non-cited violation of 10 CFR Part 50, appendix B, criterion III and associated technical specification violation of Technical Specification 3.8.1, AC Sources Operating, when the licensee failed to incorporate original circuitry design into a new digital control unit upgrade for the control building air conditioning system. This was identified when the licensee performed STP-309-0602, Division II ECCS Test, revision 056, and the chillers failed to meet the acceptance criteria and load onto the emergency diesel generator in the required time in the final safety analysis report.

Description:

On February 4, 2023, while the licensee was in mode 4 during refueling outage 22, they conducted STP-309-0602, Division II ECCS Test, revision 056, to meet TS SR 3.1.18. The purpose of the test is to simulate a LOOP in conjunction with an ECCS actuation signal and verify the de-energization of the emergency busses and load shedding of the emergency busses. Additionally, it is to verify the EDG starts on the auto start signal and energizes the emergency buses within 10 seconds and then energize the auto connected loads through the sequencing logic.

Both control building division 2 chillers, HVK-CHL1B and HVK-CHL1D, failed to sequence onto the emergency bus within the TS required time. Chiller HVK-CHL1B sequenced on the EDG in 237.1 seconds when the acceptance criteria were to sequence between 180.9 and 221.1 seconds, and HVKl-CHL1D sequenced on in 241 seconds with the same acceptance criteria. Being outside of the acceptance criteria caused the EDG to fail the surveillance and therefore be inoperable and enter the LCO, TS 3.8.1. Further investigation revealed that the division 1 control building chillers, HVK-CHL1A and HVK-CHL1C, had also failed to meet the load sequence acceptance criteria when they were tested the previous outage.

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The cause of all four chillers from both divisions failing to load onto the emergency bus in the required time was the result of a modification to upgrade the chiller controllers from analog to digital. During the upgrade under EC 00031803 completed in July 2012, a delay was introduced into the start logic that prevented them from meeting the required times. The changes made from the approved EC were implemented in 2014 and 2020 for division 1 and 2 respectively.

Corrective Actions: The licensee entered this issue into their corrective action program. The chillers sequence time was corrected under changes to the EC and new WOs.

Corrective Action References: CR-RBS-2023-01153, WO 594333(4)(5)(6), EC 95152

Performance Assessment:

Performance Deficiency: Title 10 CFR Part 50, appendix B, criterion III, Design Control, requires, that design changes, including field changes, be subject to design control measures commensurate with those applied to the original design. When implementing EC 0000311803, a design change, the licensee failed to incorporate the original design requirements for the HVK system into the EC and was therefore a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by not loading onto the emergency busses within the required times, the HVK chillers could cause a trip of the EDGs.

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, Mitigating Systems Screening Questions, the inspectors determined that this finding is of very low safety significance (Green) because the impacted SSCs maintained their operability and PRA functionality.

Cross-Cutting Aspect: P.1 - Identification: The organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. During previous executions of the ECCS surveillance, the HVK chillers failed to meet acceptance criteria and loaded onto the emergency busses outside of the required times. The licensee inappropriately used a TS bases to call the EDG operable in each instance. This was a missed opportunity to investigate why the acceptance criteria was not met and ultimately identify the error in the digital control system EC.

Enforcement:

Title 10 CFR Part 50, appendix B, criterion III, Design Control, requires, in part, that design changes, including field changes, be subject to design control measures commensurate with those applied to the original design.

SR 3.0.3 requires that if a SR is not met then the LCO is not met. Since the SR acceptance criteria was not met, the SR was failed and therefore TS 3.8.1, AC Sources Operating, was not met due to Condition D not having actions completed.

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Contrary to the above, from July 10, 2012, to April 24, 2023, the licensee performed a design change and failed to subject it to design control measures commensurate with those applied to the original design. Specifically, when implementing EC 000031803, the licensee failed to incorporate the original design requirements for the HVK system into section 3.1.25 specific circuit logic that would have temporarily bypassed the start signal for the HVK chillers until SSW flow was adequate thus preventing unnecessary trips that prevented the system from loading onto the EDG per its prescribed sequence time. This resulted in a condition that prevented a SR for TS 3.8.1 from being met and therefore inoperable for longer than time permitted by TS.

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 July 18, 2023, the inspectors presented the occupational and public radiation safety inspection results to Phil Hansett, Site Vice President, and other members of the licensee staff.

• On August 17, 2023, the inspectors presented the emergency preparedness program inspection results to Phil Hansett, Site Vice President, and other members of the licensee staff.

• On October 18, 2023, the inspectors presented the integrated inspection results to Phil Hansett, Site Vice President, and other members of the licensee staff.

• On November 13, 2023, the inspectors presented the integrated inspection results to Phil Hansett, Site Vice President, and other member of the licensee staff.

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DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Procedures EN-DC-355 Implementation of the Technical Specification Surveillance 005

Frequency Control Program

SOP-0031 Residual Heat Removal 345

SOP-0049 125VDC System 048

71111.06 Calculations PN-314 Moderate Energy Line Crack Flow Rates 0

PN-315 MELC Maximum Discharge and Suction Line Pressures for 0

RCIC, HPCS, LPCS, and RHR Fill Pumps

PN-316 Maximum Leak Rate for a Moderate Energy Line Crack 0

MELC During RHR Shutdown Cooling

PN-317 MELC-Max Flood Elevations for Moderate Energy Line 002

Cracks in Cat 1 Structures

71111.12 Corrective Action CR-RBS- 2022-05212, 2023-06233

Documents

Engineering EC 93741

Changes

Procedures EN-DC-204 Maintenance Rule Scope and Basis 008

EN-DC-205 Maintenance Rule Monitoring 009

71111.13 Corrective Action CR-RBS- 2023-05428

Documents

Procedures ADM-0096 Risk Management Program Implementation and On-Line 338

Maintenance Risk Assessment

EN-OP-119 Protected Equipment Postings 016

STP-051-4279 Containment Unit Cooler System Instrumentation Unit Cooler 11

A Functional Test

STP-051-4522 ECCS RCIC Reactor Vessel Water Level Channel Functional 14

Test (B21-692A, B21-691A)

STP-051-4524 ECCS RCIC Reactor Vessel Water Level Channel Functional 8

Test (B21-N692E, B21-691E)

71111.15 Procedures AOP-0004 Loss of Offsite Power 073

EN-OP-104 Operability Determination Process 018

EN-OP-123 Time Critical Action Program Standard 007

71111.24 Corrective Action CR-RBS- 2023-04358, 2023-06044, 2023-06570

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Inspection Type Designation Description or Title Revision or

Procedure Date

Documents

71114.02 Miscellaneous River Bend Station Alert and Notification System Design 01/04/2013

Report

Procedures EPP-2-701 Prompt Notification System Maintenance and Testing 34

71114.03 Miscellaneous ERO Notification System Test Results 03/23/2022

ERO Notification System Test Results 09/20/2022

ERO Notification System Test Results 06/21/2023

ERO Notification System Test Results 12/17/2021

River Bend Station On-Shift Staffing Analysis Final Report 1

Procedures EIP-2-006 Notifications 49

EN-EP-310 Emergency Response Organization Notification System 11

EPP-2-502 Emergency Communications Equipment Testing 29

71114.04 Miscellaneous 10CFR50.54(Q)(2) Review for EN-EP-202 Revision 3 07/26/2021

10CFR50.54(Q)(2) Review for EIP-2-024 Revision 27 01/12/2023

10CFR50.54(Q)(3) Screening for PR-C-422 09/12/2022

10CFR50.54(Q)(2) Review for EIP-2-016 Revision 32 11/01/2021

Procedures EN-EP-305 Emergency Planning 10CFR50.54(q) Review Program 8

71114.05 Corrective Action CR-RBS- 2021-04918, 2021-05407, 2021-05963, 2021-06899, 2022-

Documents 02007, 2022-02599, 2022-03308, 2022-03409, 2022-03410,

22-03629, 2022-05741, 2022-05992, 2022-06138, 2022-

2420, 2022-06185, 2023-00407, 2023-02200, 2023-02207,

21-03346, 2021-03355, 2021-03512, 2021-03515

Corrective Action CR-RBS- 2023-06592, 2023-06596

Documents

Resulting from

Inspection

Miscellaneous ERO Team B Site Drill Report 10/13/2021 11/12/2021

ERO Team B Site Drill Report 11/10/2021 03/01/2022

ERO Team C Site Drill Report 02/23/2022 03/24/2022

22 NRC IPX Exercise Drill Report 11/14/2022

Our Lady of the Lake MS-1 Drill Report 03/16/2022

Medical Emergency Drill On-Site 12/13/2022

Team B EOF Focused Drill Report 10/05/2022 10/06/2022

2-36

Inspection Type Designation Description or Title Revision or

Procedure Date

Team D EOF Focused Drill Report 10/12/2022 10/24/2022

22 EMPE Drill Report 06/15/2022 07/14/2022

22 Accountability Drill Report 12/22/2022 01/10/2023

HP Drill Report 06/22/2023 07/18/2023

Emergency Plan 48

2nd Quarter Team C Focused Drill Report 06/28/2023 07/12/2023

QA-7-2021-RBS- Emergency Preparedness Quality Assurance Audit Report 06/28/2020

QA-7-2023-RBS- Emergency Preparedness Quality Assurance Audit Report 04/28/2022

QS-2022-RBS- Emergency Preparedness Quality Assurance Surveillance 04/12/2022

2 Report

Procedures ADM-0060 First Aid Team Emergencies 16

EIP-2-001 Classification of Emergencies 30

EIP-2-002 Classification Actions 37

EIP-2-007 Protective Action Recommendation Guidelines 28

EN-EP-306 Drills and Exercises 11

EN-EP-609 Emergency Operations Facility (EOF) Operations 7

EN-EP-610 Technical Support Center (TSC) Operations 8

EN-EP-611 Operations Support Center Operations 8

EN-TQ-110 Emergency Response Organization Training 15

EPP-2-501 Emergency Facilities and Equipment Readiness 18

Work Orders WO-545130

71124.05 Corrective Action CR-RBS- 2021-03740, 2021-04317, 2021-05817, 2021-06840, 2022-

Documents 02036, 2022-04484, 2022-4618, 2022-4860, 2022-06323,

22-06714, 2022-0621, 2023-04854

Corrective Action CR-RBS- 2023-05769, 2023-05771

Documents

Resulting from

Inspection

Miscellaneous Table of Routine Source Check Sources, Assay Date, and 06/12/2023

Current Activity

Procedures EN-RP-302 Operation of Radiation Protection Instrumentation 6

2-37

Inspection Type Designation Description or Title Revision or

Procedure Date

EN-RP-306 Calibration and Operation of the Eberline PM-7 3

EN-RP-307 Operation and Calibration of the Eberline Personal 2

Contamination Monitors

EN-RP-308 Operation and Calibration of Gamma Scintillation Tool 9

Monitors

EN-RP-312 Operation and Calibration of the Canberra GEM-5 4

EN-RP-313 Operation and Calibration of the ARGOS-5AB Personnel 3

Contamination Monitor

EN-RP-315 Operation and Calibration of the CRONOS Contamination 3

Monitor

EN-RP-317-05 Calibration of Extendable Dose Rate Instruments 1

EN-RP-317-09 Calibration of Dosimeters 5

EN-RP-317-10 Calibration of Portable Dose Rate Instruments 3

RHP-0106 Calibration of the Canberra Fastscan and Accuscan II Whole 4

Body Counters

STP-511-4237 Main Plant Exhaust Duct Monitoring System Flow Rate 305

Monitor Channel Calibration RMS-FEX125, RMS-FEY125

Radiation RBS-2306-00116 3119 TB 95 Chemistry Sample Area 06/07/2023

Surveys RBS-2306-00400 3000 Turbine Building 67-foot Elevation 06/26/2023

Work Orders Work Order (WO) 52769324-01, 52948224-01, 00565523-01, 00585136-01,

00596070-01,

71124.08 Corrective Action CR-RBS- 2022-00723, 2022-04574, 2023-01499, 2023-02069, 2023-

Documents 02738

Corrective Action CR-RBS- 2023-05914, 2023-05940, 2023-05729, 2023-05730, 2023-

Documents 05732, 2023-05735, 2023-05930, 2023-05931, 2023-05933

Resulting from

Inspection

Miscellaneous Dry active waste, waste stream analysis 04/17/2023

Reactor water cleanup system waste stream analysis 08/09/2022

Radioactive source inventory list 06/27/2023

FCBT-MPC- Shipping training for radwaste workers 1

HAZSEC

FCBT-MPC-TSP Shipping security training for radwaste workers 3

Procedures EN-RP-121 Radioactive Material Control 18

2-38

Inspection Type Designation Description or Title Revision or

Procedure Date

EN-RW-102 Radioactive Shipping Procedure 20

EN-RW-104 Scaling Factors 14

EN-RW-105 Process Control Program 5

EN-RW-106 Integrated Transportation Security Plan 7

SOP-0112 Solid Radwaste Processing 22

SOP-012 Solid Radwaste Collection 8

Self-Assessments LO-RLO-2020- RP Pre-NRC focused self-assessment plan 02/16/2021

00018

QA-14/15-2021- Combined Radiation Protection (RP) and Radwaste (RW) 09/16/2021

RBS-01 audit

Shipping Records RBS-2022-031 Shipping package for shipment number RBS-2022-031

RBS-2022-034 Shipping package for shipment number RBS-2022-034

RBS-2022-059 Shipping package for shipment number RBS-2022-059

RBS-2023-040 Shipping package for shipment number RBS-2023-040

RBS-2023-041 Shipping package for shipment number RBS-2023-041

71151 Miscellaneous Alert and Notification System Test Data 4Q2022 - 2Q2023

Drill and Exercise Performance Data 4Q2022 - 2Q2023

Emergency Response Organization Drill Participation Data

4Q2022 - 2Q2023

71152S Procedures ADM-0073 Temporary Services and Equipment 307

EDS-ME-002 Control of Loose Items 002

EN-OP-115 Conduct of Operations 031

EN-OP-115-09 Maintaining the Station Narrative Log 004

2-39