IR 05000269/2023004
| ML24038A367 | |
| Person / Time | |
|---|---|
| Site: | Oconee, 07200040  |
| Issue date: | 02/13/2024 |
| From: | Eric Stamm NRC/RGN-II/DRP/RPB1 |
| To: | Snider S Duke Energy Carolinas |
| References | |
| IR 2023002 | |
| Download: ML24038A367 (33) | |
Text
SUBJECT:
OCONEE NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000269/2023004, 05000270/2023004, AND 05000287/2023004; AND INSPECTION REPORT 07200040/2023002
Dear Steven Snider:
On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Oconee Nuclear Station. On February 1, 2024, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. These findings involved violations of NRC requirements; one of these violations was determined to be Severity Level IV. Additionally, one Severity Level IV violation associated with a previously identified finding is documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Oconee Nuclear Station.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Oconee Nuclear Station.February 13, 2024 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Eric J. Stamm, Chief Reactor Projects Branch 1 Division of Reactor Projects Docket Nos. 05000269, 05000270, 05000287 07200040 License Nos. DPR-38, DPR-47, and DPR-55
Enclosure:
As stated
Inspection Report
Docket Numbers: 05000269, 05000270, and 05000287; 07200040
License Numbers: DPR-38, DPR-47, and DPR-55
Report Numbers: 05000269/2023004, 05000270/2023004, and 05000287/2023004; 07200040/2023002
Enterprise Identifier: I-2023-004-0021; I-2023-002-0076
Licensee: Duke Energy Carolinas, LLC
Facility: Oconee Nuclear Station
Location: Seneca, South Carolina
Inspection Dates: October 1, 2023, to December 31, 2023
Inspectors: J. Nadel, Senior Resident Inspector E. Robinson, Resident Inspector N. Smalley, Resident Inspector J. Bell, Senior Health Physicist P. Cooper, Senior Reactor Inspector J. Diaz-Velez, Senior Health Physicist W. Erling, Reactor Inspector D. Neal, Health Physicist A. Nielsen, Senior Health Physicist J. Parent, Resident Inspector A. Ruh, Reactor Inspector M. Schwieg, Senior Reactor Inspector
Approved By: Eric J. Stamm, Chief Reactor Projects Branch 1 Division of Reactor Projects
Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Oconee Nuclear Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Adequately Implement Containment Materiel Controls Resulting in Unborated Water Left in Containment Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.14] - 71111.20 NCV 05000270/2023004-01 Conservative Open/Closed Bias The inspectors identified a Green finding and associated non-cited violation (NCV) of technical specification (TS) 5.4.1.a, Procedures, when the licensee failed to adequately implement containment materiel controls in Unit 2 containment. Specifically, when starting up from a refueling outage, the reactor vessel head stand water shields were discovered to be full of unborated water on November 19, 2023.
Failure to Log a Technical Specification Entry and Report a Condition Prohibited by Technical Specifications for a Unit 3 Low Pressure Service Water System Leak in Containment Cornerstone Significance/Severity Cross-Cutting Report Aspect Section Barrier Integrity Green [P.2] - 71152A Severity Level IV Evaluation NCV 05000287/2023004-02 Open/Closed The inspectors identified a Green finding and associated Severity Level IV NCV of TS 5.4.1.a,
Procedures, and 10 CFR 50.73(a)(2)(i)(B), when the licensee failed to make operator narrative log entries for the entry and exit from TS limiting condition for operation (LCO) 3.0.3, for a low pressure service water piping leak in Unit 3 containment and subsequent failure to issue a licensee event report (LER) within 60 days of a condition prohibited by TS associated with the required TS LCO 3.0.3 entry.
Failure to Report a Condition That Could Have Prevented Fulfillment of a Safety Function Associated with Online Reactor Building Cooling Unit Cleaning Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000270/2023004-03 Open/Closed The inspectors identified a Severity Level IV NCV of 10 CFR 50.73(a)(2)(v) when the licensee failed to issue an LER within 60 days of a condition that could have prevented fulfillment of a safety function. Specifically, on August 25, 2023, the licensee received an NRC comprehensive engineering team inspection report, which documented a condition that could have prevented fulfillment of a safety function associated with a procedure violation for the online cleaning of reactor building cooling units (RBCUs), but no NRC notifications were made.
Additional Tracking Items
None
PLANT STATUS
Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection period except on October 18, 2023. The unit was reduced to 25 percent RTP for a balance shot to correct vibrations on the 1B2 reactor coolant pump. The unit was returned to 100 percent RTP on the same day.
Unit 2 began the inspection period at or near 100 percent RTP. On October 27, 2023, Unit 2 was shut down for a scheduled refueling outage (O2R31). On November 23, 2023, Unit 2 was returned to 100 percent RTP and operated at or near 100 percent RTP for the remainder of the inspection period.
Unit 3 operated at or near 100 percent RTP for the entire inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2 electrical distribution system during the O2R31 refueling outage on November 14, 2023
- (2) Unit 2 emergency feed water system on December 20, 2023
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the Unit 2 low pressure injection (LPI) system during the O2R31 refueling outage between October 30 and November 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) Fire zone 95: Unit 1 equipment room on October 2, 2023
- (2) Fire zone 77: Unit 3 auxiliary building 200 level hallway on October 2, 2023
- (3) Fire zone 33: Unit 2 4160V switchgear on October 10, 2023
- (4) Fire zone 123: Unit 2 containment on November 9, 2023
- (5) Fire zone 103: Unit 2 east penetration room on November 28, 2023
===71111.08P - Inservice Inspection Activities (PWR)
The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from October 30, 2023, to November 9, 2023.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===
The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:
- (1) Ultrasonic testing of reactor vessel internals
- flow distributor bolts (quantity: 96) (reviewed)
- baffle-to-former bolts (quantity: 108) (reviewed)
- lower core barrel bolts (quantity: 108) (reviewed)
PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection
Activities (IP Section 03.02) (1 Sample)
The inspectors verified that the license conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately:
- (1) Visual examination (VT-2) of reactor vessel head bare metal visual (observed and reviewed)
PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)
The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:
(1)
- boric acid evaluation on 2A LPI pump piping 1-HPI-IV-0137 and 2LWD-370 threaded connections
- independent boric acid containment walkdowns
PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04) (1 Sample)
The inspectors verified that the licensee is monitoring the steam generator tube integrity appropriately through a review of the following examinations:
(1)
- steam generator 2A - eddy current testing tubes
- steam generator 2B - eddy current testing tubes
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03) (1 Sample)
The licensee completed the annual requalification operating examinations and biennial written examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), "Requalification Requirements," of the NRC's "Operator's Licenses." The inspector performed an in-office review of the overall pass/fail results of the individual operating examinations, the crew simulator operating examinations, and the biennial written examinations in accordance with IP 71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance." These results were compared to the thresholds established in Section 3.03, "Requalification Examination Results," of IP 71111.11.
- (1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam administered on March 23, 2023, and the biennial written examinations completed on March 24, 2023.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during Unit 1 power reduction from 100 percent to 25 percent for a 1B2 reactor coolant pump balance shot on October 18, 2023.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated simulator just in time training for pressurizer cooldown using lesson plan OP-OC-16JT-06 on October 23, 2023.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (3 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Nuclear condition report (NCR) 2477210, 3A1 reactor coolant pump upper bearing oil cooler low pressure service water pipe leak inside containment, on June 25, 2023
- (2) NCR 2478801, 7C and 8C Lee combustion turbines failed to run, on July 10, 2023
- (3) NCR 2490763, primary instrument air compressor failed to load, on October 17, 2023
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 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) Unit 3 green risk during low pressure injection pump maintenance, on October 12, 2023
- (2) Unit 1 green risk due to low pressure injection, condenser circulating water, and Keowee hydro unit maintenance, on October 25, 2023
- (3) Unit 2 yellow risk during draining of the reactor coolant system loops, on October 30, 3023
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (9 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) NCR 2459768, gas void identified at 1LP-208 during monthly ultrasonic testing and venting procedure
- (2) NCR 2491324, through wall active leak on 3C essential siphon vacuum (ESV) pump air/water separator
- (3) NCR 2492398, ESV float valve vent line is not currently heat traced
- (5) NCR 2493300, 2LPSW-1236 pipe pitting discovered under allowable thickness
- (7) NCR 2495280, Unit 2 reactor vessel head stand water shields found full of water in mode 4
- (8) NCR 2477210, Unit 3 low pressure service water leak on outlet of the 3A1 reactor coolant pump upper bearing cooler
- (9) NCR 02492966, 2HP-103 and 2HP-993 reach rods not operational (cannot be positioned)
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated refueling outage O2R31 activities from October 27, 2023, to November 22, 2023.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (5 Samples)
- (1) PT/0/A/0400/016, "SSF HVAC System Flow Test," following filter and belt replacement on air handling unit 0-42, on October 14, 2023, work order (WO)
===20629409
- (2) Functional test of Unit 2 LPI suction header relief valve following replacement, on October 26, 2023, WO 20460059
- (3) Replacement of deformed 'Y' and 'Z' phase current transformers on B2T-4, on November 6, 2023, WO 20632933
- (4) LPSW-1236 post-maintenance test plan in accordance with EC 420875 following piping and valve replacement, on November 13, 2023, WO 20547398
- (5) Retest of 2B2 reactor coolant pump following motor replacement and associated maintenance, on November 15, 2023, WO 20583997
Surveillance Testing (IP Section 03.01)===
- (1) PT/0/A/0400/015, "SSF Submersible Pump Test," at CCW intake structure, on October 3, 2023
- (2) MP/0/B/2002/001, "Inspection of Keowee Underground Cable Trench Drainage System," on October 23 and 24, 2023
- (3) PT/2/A/0261/020, "Essential Condenser Circulating Water System Test," on October 29, 2023
- (4) PT/2/A/0251/069, "Unit 2 LPI Cooler Test," on October 29, 2023
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) Observed leak rate testing on penetrations 60 and 61, on November 9, 2023
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
- (1) Control of radioactive material in the Unit 1 & 2 spent fuel pool.
- (2) Licensee surveys of potentially contaminated material leaving the radiologically controlled area (RCA).
Radiological Hazards Control and Work Coverage (IP Section 03.04) (3 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Unit 1 at-power entry
- (2) Unit 2 head lift
- (3) Unit 2 plenum move
High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (4 Samples)
The inspectors evaluated licensee controls of the following high radiation areas and very high radiation areas:
- (1) Unit 2 vessel head stand
- (2) Unit 2 under vessel entry
- (3) Unit 2 reactor coolant bleed holdup tank room
- (4) Unit 1 CRD filter room
Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Permanent Ventilation Systems (IP Section 03.01) (1 Sample)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
- (1) Unit 3 main control room ventilation filtration system
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) Reactor vessel head stand High Efficiency Particulate Air (HEPA) unit
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.
External Dosimetry (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.
Internal Dosimetry (IP Section 03.03) (3 Samples)
The inspectors evaluated the following internal dose assessments:
- (1) Worker #1, internal dose assessment performed on April 27, 2022
- (2) Worker #2, internal dose assessment performed on April 27, 2022
- (3) Worker #3, internal dose assessment performed on April 27, 2022
Special Dosimetric Situations (IP Section 03.04) (1 Sample)
The inspectors evaluated the following special dosimetric situations:
- (1) The inspectors reviewed dose assessments for two Declared Pregnant Workers (DPW), declared on May 11, 2022, and August 3, 2023.
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) 0RIA-33, RWF Liquid Radwaste Radiation Monitor
- (2) 1RIA-12, Auxiliary Building 2nd floor, Area Radiation Monitor
- (3) 3RIA-12, Auxiliary Building 2nd floor, Area Radiation Monitor
- (4) 1RIA-50, Auxiliary Building 2nd floor, Area Radiation Monitor
- (5) 1/2RIA-6, SFP Area Radiation Monitor
- (6) 3RIA-13, Auxiliary Building, Area Radiation Monitor
- (7) Argos Zeus, RCA exit, Whole Body Contamination Monitor
- (8) RIA-RT-0045, RWF, Noble Gas Monitor
Calibration and Testing Program (IP Section 03.02) (13 Samples)
The inspectors evaluated the calibration and testing of the following radiation detection instruments:
- (1) Ludlum 177, Serial Number: 312618, calibrated August 28, 2023
- (2) Ludlum 177, Serial Number: 312670, calibrated August 31, 2023
- (3) Ludlum 9-3, Serial Number: 288760, calibrated August 14, 2023
- (4) Ludlum 9-3, Serial Number: 279930, calibrated August 14, 2023
- (5) Canberra iCAM, Serial Number: 6088, calibrated July 17, 2023
- (6) Argos 5AB Zeus, Serial Number: 2105-096-AR050AB calibrated July 20, 2023
- (7) Argos 5AB Zeus, Serial Number: 2105-095-AR050AB calibrated July 20, 2023
- (8) 3RIA-1, Unit 3 Control Room Radiation Monitor calibrated February 17, 2011, and June 30, 2021
- (9) ONSF1, Admin Building, Room 0109, Apex In-vivo Whole Body Counter, calibrated September 23, 2023
- (10) Rotem-Telepole, Serial Number: 6604-068, calibrated September 19, 2023
- (11) Rotem-Telepole, Serial Number: 6604-089, calibrated September 19, 2023
- (12) Rotem-AMP 100, Serial Number: 5004-047, calibrated January 27, 2023
- (13) Rotem-AMP 100, Serial Number: 5002-034, calibrated March 2, 2023
Effluent Monitoring Calibration and Testing Program Sample (IP Section 03.03) (3 Samples)
The inspectors evaluated the calibration and maintenance of the following radioactive effluent monitoring and measurement instrumentation:
- (1) RWF RIA-RT-0033, Liquid Radwaste Radiation Monitor, calibrated March 10, 2022, and April 13, 2023
- (2) U1-RIA-45, Unit 1 Vent Gas, normal range calibrated June 13, 2022, and July 3, 2023
- (3) U1-RIA-46, Unit 1 Vent Gas, high range calibrated June 21, 2018, and October 19,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15)===
- (1) May 1, 2022, through October 3, 2023
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program for potential adverse trends in complacency and procedure use and adherence that might be indicative of a more significant safety issue.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
===60853 - Onsite Fabrication of Components and Construction of an Independent Spent Fuel Storage Installation
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 IMC 2690, Inspection Program for Storage of Spent Reactor Fuel and Reactor-Related Greater-than-Class C Waste at Independent Spent Fuel Storage Installations (ISFSI)and for 10 CFR Part 71 Transportation Packagings." 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.
Onsite Fabrication of Components and Construction of an Independent Spent Fuel Storage Installation===
- (1) On August 17, 2023, to October 17, 2023, the inspector conducted a review of licensee and vendor activities in preparation for the construction of a concrete storage pad. The inspectors verified that activities related to the ISFSI have been properly incorporated into the existing licensee programs by reviewing the following activities:
1. ISFSI pad design: The inspector verified that the licensee completed evaluations which establish that the ISFSI storage pads and areas have been designed to adequately support the static and dynamic loads of the stored dry storage systems (DSS), considering potential amplification of earthquakes through soil-structure interaction, and soil liquefaction potential or other soil instability due to vibratory ground motion.
2. ISFSI pad construction: The inspector verified the ISFSI storage pad was constructed in accordance with the DSS Certificate of Compliance and reviewed licensee controls that ensured no adverse impact on-site operations or TS by performing the following activities; a. The inspector walked down the construction area of the ISFSI pad to verify the formwork and rebar placement complied to licensee-approved drawings, and specifications.
b. The inspectors observed the placement of the ISFSI slab, and observed tests for concrete slump, air content, temperature measurements, and the collection/preparation of cylinder samples for compression tests, to verify that the work was implemented in accordance with the approved specifications and procedures, c.
The inspector verified that the pad was being finished according to approved specifications and code requirements.
d. Following completion of the 7-day and 28-day compression tests by the independent laboratory, the inspector reviewed the results to verify that the acceptance criteria were met.
INSPECTION RESULTS
Failure to Adequately Implement Containment Materiel Controls Resulting in Unborated Water Left in Containment Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.14] - 71111.20 NCV 05000270/2023004-01 Conservative Open/Closed Bias The inspectors identified a Green finding and associated non-cited violation (NCV) of technical specification (TS) 5.4.1.a, Procedures, when the licensee failed to adequately implement containment materiel controls in Unit 2 containment. Specifically, when starting up from a refueling outage, the reactor vessel head stand water shields were discovered to be full of unborated water on November 19, 2023.
Description:
The Emergency Core Cooling System (ECCS) is designed to cool the reactor core and provide shutdown capability following initiation of a Loss of Coolant Accident (LOCA). To ensure design system capability for each new operating cycle after a refueling outage, a core reload design check is completed by the licensee. Part of the core reload design verification process is to ensure that the mixed reactor building emergency sump (RBES) fluid boron concentration following a LOCA is greater than required to keep the reactor shutdown (subcritical) in the recirculation phase of a LOCA event initiated from Hot Full Power conditions. The analysis for this check assumes the mixed sump fluid is made of borated water from the Reactor Coolant System (RCS), the Core Flood Tanks (CFT), and the Borated Water Storage Tank (BWST) at their respective minimum allowed boron concentration as specified in the approved Core Operating Limits Report (COLR) for that cycle.
On November 19, 2023, after the licensee had completed containment closeout inspection activities on Unit 2 in preparation to enter Mode 4, the inspectors entered containment and, based on visual indications, questioned if the reactor vessel head stand water shields were still full of water, contrary to licensee procedures. Upon notification of the concern by the inspectors, the licensee entered containment and verified that the water shields did in fact contain water. These water shields are filled at the beginning of the outage with unborated water as a dose reduction tool to support refueling outage activities. The total volume of the seven water shields is 420 cubic feet, which equates to about 3,141 gallons of water. The purpose of MP/0/A/3005/012, Revision 20, Containment Inspection/Closeout Procedure, is to provide for complete and consistent closeout and inspection of the containment prior to entry to Mode 4. This procedure is used to inspect containment for housekeeping, materiel condition, and foreign material exclusion to establish RBES operability. This procedure implements OP/2/A/1102/028, Revision 21, Reactor Building Tour and Containment Materiel Controls, which in part provides guidance for containment materiel controls in Modes 1-4 as well as guidance for permanent storage of materiel in the reactor building during the operating cycle. OP/2/A/1102/028, Enclosure 4.22, contains a list of items that are approved to be permanently stored in containment. Although the specified reactor vessel head stand water shields are authorized to be permanently stored in containment, there is a clarifying note that requires that the water shields shall be drained prior to entering Mode 4. During a LOCA event, this water could be released from the shields to the reactor building basement and therefore interact with the mixed RBES fluid. The potential additional unborated water introduced to the RBES could reduce the margin of subcriticality and potentially challenge the ECCS safety function to keep the core subcritical in the post-LOCA recirculation phase of ECCS operation.
The licensee entered this issue into its corrective action program (CAP) as NCR 2495280 and performed an operability determination (OD). The initial OD included input from fleet safety analysis personnel and design engineering based on the current condition of the plant in Mode 4. After further interaction with the resident inspector staff, the OD was updated to include an analysis of the condition for the entire operating cycle. The OD concluded that, although the additional unborated water from the undrained water shields impacted the margin to the post-LOCA subcriticality design check, operability of the ECCS safety function to keep the core subcritical in the post-LOCA recirculation phase was not impacted and no reload core design goals would have been missed. The licensee drained the water shields later that night on November 19, 2023.
The licensee completed a cause evaluation checklist to determine the cause of the event as well as identify further appropriate corrective actions as necessary. A Work Order (WO) was identified that was scheduled to be completed at the end of the outage that would have drained the reactor vessel head stand water shields. However, this WO was closed out with no work performed due to a misunderstanding by maintenance personnel of the implementation of a time-savings effort to leave the water shields in containment permanently. Per an evaluation performed under NCR 2302691, the water shields were approved to be left in containment during the operating cycle, but only in a fully drained condition. This condition was not identified by the maintenance personnel completing the containment closeout procedure, MP/0/A/3005/012. Based on interviews and statements from licensee personnel, multiple opportunities existed for rigorous communication and validation of assumptions to occur prior to the decision not to drain the water shields.
Corrective Actions: The licensee entered this violation into its CAP, performed an operability determination, and drained the water shields.
Corrective Action References: NCR 2495280
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to adequately implement containment materiel controls prior to startup in accordance with licensee procedure MP/0/A/3005/012 was a performance deficiency. Specifically, containment materiel controls were not adequately implemented when unborated water from the reactor vessel head stand water shields was not drained during containment close out.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the additional unborated water potentially introduced to the RBES during a LOCA event could reduce the mixed sump boron concentration and therefore reduce the margin of subcriticality and potentially impact the ECCS safety function to keep the core subcritical in the post-LOCA recirculation phase of ECCS operation.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Per IMC 0609, Attachment 4, Initial Characterization of Findings, the finding is associated with the Fuel Cladding Integrity aspect of the Barrier Integrity Cornerstone. Using Exhibit 3, Barrier Integrity Screening Questions, inspectors determined the finding was of very low safety significance (Green) because the finding did not involve control manipulations that unintentionally added positive reactivity, did not result in mismanagement of reactivity by operators, did not result in the mismanagement of foreign material exclusion or the reactor coolant chemistry control program, and did not result from fuel handling errors.
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. In this case, the licensee relied solely on informal communications between maintenance personnel when presented with a decision not to drain the reactor vessel head stand water shields and failed to validate the required condition of water shields per an approved procedure.
Enforcement:
Violation: TS 5.4.1.a, Procedures, requires, in part, that written procedures be established, implemented, and maintained covering activities related to procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978. Regulatory Guide 1.33, Section 3(f),
Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems, requires procedures addressing Containment, which was partially implemented by MP/0/A/3005/012, Revision 20. MP/0/A/3005/012, sections 11.1.5, 11.2.8, 11.3.5, 12.1, and 12.2 required, in part, that personnel ensure materiel is controlled in containment per OP/2/A/1102/028, Revision 21, during containment closeout prior to startup. Contrary to the above, the licensee failed to ensure materiel was controlled in containment per OP/2/A/1102/028 during containment closeout when the licensee failed to drain the reactor vessel head stand water shields prior to entry to Mode 4 on November 19, 2023.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Log a Technical Specification Entry and Report a Condition Prohibited by Technical Specifications for a Unit 3 Low Pressure Service Water System Leak in Containment Cornerstone Significance/Severity Cross-Cutting Report Aspect Section Barrier Green [P.2] - 71152A Integrity Severity Level IV Evaluation NCV 05000287/2023004-02 Open/Closed The inspectors identified a Green finding and associated Severity Level IV NCV of TS 5.4.1.a, Procedures, and 10 CFR 50.73(a)(2)(i)(B), when the licensee failed to make operator narrative log entries for the entry and exit from TS limiting condition for operation (LCO) 3.0.3, for a low pressure service water piping leak in Unit 3 containment and subsequent failure to issue a licensee event report (LER) within 60 days of a condition prohibited by TS associated with the required TS LCO 3.0.3 entry.
Description:
The low pressure service water (LPSW) system provides cooling water to various emergency and non-emergency systems throughout the plant. One of the non-emergency systems served by LPSW is the reactor coolant pump (RCP) upper motor bearing oil coolers. On June 22, 2023, after completing a routine makeup to the reactor coolant system (RCS), operators in the Unit 3 control room identified that the reactor building normal sump showed an increased inleakage rate. The licensee analyzed all system inputs into the sump and performed an RCS leak rate calculation. RCS leakage was ruled out based on the results of the calculation and past trends. Unit 3 was reduced to approximately 20 percent power on June 24, 2023, to allow for leak inspections in the steam generator cavities inside containment. On June 24, 2023, the leak was identified as coming from 0.25-inch diameter LPSW system piping associated with the 3A1 RCP upper motor bearing oil cooler, upstream of normally closed instrument valve 3LPS-IV-0081, which had completely sheared off.
Subsequent extent of condition walkdowns identified an additional active leak near instrument valve 3LPS-IV-0082. The Unit entered mode 3 to allow repairs to the piping on June 25, 2023, and returned to 100 percent power on June 27, 2023, following successful repairs. The residents noted that the failed section of piping was credited within the design bases as a closed system inside containment and, along with two single active containment isolation valves, provided the redundant isolation barrier for two separate containment penetrations that support the containment safety function of limiting leakage of fission product radioactivity during an accident. For more information on the piping failure that caused the leak, see NRC finding 05000287/2023003-01 (ADAMS Accession No. ML23310A171).
During the identification and repair of the LPSW leak, the licensee did not enter any TS LCO actions related to the degraded condition. When inspectors questioned this at the time, the licensee reviewed their TS, held several internal meetings, and consulted with fleet resources to discuss the applicability of TS to the degraded condition. At the time, the licensee determined that no TS entry was required.
The resident inspectors determined that reasonable and adequate bases existed to conclude that TS 3.6.3 does apply to a degraded closed system inside containment.
The TS 3.6.3 Bases at Oconee state, in part:
The containment isolation valves form part of the containment pressure boundary and provide a means for fluid penetrations not serving accident consequence limiting systems to be provided with two isolation barriers that are closed on an automatic isolation signal. These isolation devices consist of either passive devices or active (automatic) devices. Manual valves, non-automatic power operated valves in their closed position, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), blind flanges, and closed systems are considered passive devices.
The normally closed isolation valves are considered OPERABLE whenclosed systems are intact.
This LCO provides assurance that containment isolation valveswill perform their designated functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.
Condition C is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve in closed systems. This Note is necessary since this Condition is written to specifically address those penetration flow paths in a closed system.
The inspectors concluded, based on the Oconee TS as explained in the bases, that in order for LCO 3.6.3 to be met, a closed system that is credited as one of the containment isolation barriers must be intact. For a closed system that is part of the isolation boundary and has a single isolation valve at each penetration path, TS 3.6.3 applies to both the valve and the closed system and should be entered any time the closed system is breached. Therefore, the June 2023 LPSW piping failure met the conditions for TS LCO 3.6.3 to apply. Since no action for a breached or inoperable closed system exists in TS 3.6.3, entry into TS 3.0.3 is required.
This is because, as discussed in the bases, TS 3.6.3 actions are written based on an inherent assumption that the closed systems that form the inside containment boundary remain intact.
TS usage requirements in accordance with TS 3.0.3 require entry into TS 3.0.3 if an associated LCO, in this case TS LCO 3.6.3, is not met and there are no actions provided within that LCO. The time of discovery that the closed system was no longer intact was determined to be June 24, 2023, at 2148.
Oconee QA-1 calculations OSC-7005, Revision 0, and OSC-8966, Revision 0, both titled Allowable LPSW Leakage for Containment Integrity, indicate that leakage from LPSW, at the containment elevation where the pipe failed, which is greater than approximately 0.5 gallons per minute is enough to exceed the containment atmospheric volume criteria (La)limit required for containment operability under TS 3.6.1. The observed leakage from the broken LPSW pipe was calculated at over four gallons per minute (gpm).
Inspectors noted that Duke procedure AD-OP-ALL-0112, Operations Log Keeping and Chart Recorders, Revision 1, sections 5.1.1.9 and 5.1.3.3, require the narrative logs be updated with LCO entry and exit for abnormal plant alignments and with actions taken to comply with Technical Specifications and other regulatory requirements. However, no log entries were made regarding applicable TS LCOs associated with the LPSW piping failure in June 2023.
It was noted that based on reactor building normal sump (RBNS) inleakage rates during the event, the leakage exceeded the containment operability LPSW leakage criteria (~0.5 gpm)at approximately 1240 on June 23, 2023. Similarly, the reduction of the leak rate to less than 0.5 gpm through the closure of automatic valves 3LPSW-7 and 2LPSW-8 at 0834 on June 25, 2023, would have potentially allowed exit from applicable TS LCO 3.0.3 required actions.
The following is the relevant timeline of events:
Date Time Event 06/22/2023 - 2250 - Unit 3 nightshift operators completed a makeup and shortly afterwards noticed RBNS makeup rate was higher than normal, kicking off leak-hunt activities 06/23/2023 - 1240 - The calculated leak rate increased to over approximately 0.5 gpm, exceeding the OSC-7005/8966 LPSW leakage limits (For reporting purposes, firm evidence existed)06/24/2023 - 2048 - Reactor power reached 20% to allow access to the steam generator cavities for leak-hunt 06/24/2023 - 2148 - Containment entry team reported the leak location to the control room after making entry into the unit 3 A cavity inside containment. (For TS logging purposes, time of discovery)06/25/2023 - 0540 - Unit 3 entered Mode 3 for leak repair activities 06/25/2023 - 0834 - The clearance for leak repair activities reduced leakage from the break in the RCP 3A1 upper motor bearing LPSW piping through the closure of automatic isolation valves 3LPSW-7 and 3LPSW-8 06/25/2023 - 2300 - Pipe repair was completed, and the closed system was restored to intact 06/26/2023 - 1629 - Unit 3 entered mode 2 on startup
Technical Specification LCO 3.0.3 requires, in part:
When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable.
Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:
a. MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; b. MODE 4 within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />; and c. MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
Through a review of the above events, the residents concluded that LCO 3.0.3 applied to the condition of the degraded LPSW closed system inside containment at the time firm evidence existed of a leak exceeding the 0.5 gpm criteria, which is 1240 on June 23, 2023. Therefore, TS LCO 3.0.3.a, b, and c required actions times were not met prior to the leak reduction actions taken at 0834 on June 25, 2023, which is approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> after TS 3.0.3 entry would have been required. This represents a condition prohibited by TS in accordance with 10 CFR 50.73(a)(2)(i)(B). It should be noted that the leakage was greatly reduced, but still present, after the 3LPSW-7/8 valve closures occurred. The licensee did not formally evaluate the operability of the closed system barrier in this configuration in light of TS bases assumptions that rely on the closed system remaining intact.
On July 27, 2023, the licensee documented these issues in NCR 2477611. The licensees review and determination of the reportability of the events was tracked by NCR assignment 2477611-1. This assignment was not completed until February 5, 2024; concluding that the degraded condition is reportable under 50.73(a)(2)(i)(B) for the reasons above. The licensee plans to make the required report by the end of February 2024.
Corrective Actions: The licensee wrote NCR 2503209, which acknowledged the applicability of TS LCO 3.6.3, LCO 3.0.3, and reporting condition 10 CFR 50.73(a)(2)(i)(B) to the degraded LPSW piping. A standing instruction that provided guidance to operators in the case of failed closed systems inside containment was also updated to align with the LCO 3.6.3/3.0.3 position.
Corrective Action References: NCRs 2477145, 2477210, 2477611, 2489821
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to track applicable technical specification LCOs as required by section 5.1.1.9 and 5.1.3.3 of AD-OP-ALL-0112, was a performance deficiency. This requirement was not satisfied because operators failed to identify and track LCO 3.0.3 after identification of the degraded LPSW piping in containment. The NRC determined this violation was associated with a previously documented finding assessed using the significance determination process.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, this failure was indicative of a weakness with the licensees implementation of Technical Specifications, the design bases of containment and containment penetrations, and the application of the single failure criterion. This weakness resulted in misapplication of TS and could have led to missed TS-required actions if the vulnerability remained unrecognized.
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 3, Barrier Integrity Screening Questions, inspectors determined the finding was of very low safety significance (Green) because the performance deficiency was strictly an administrative requirement to track inoperable equipment and did not result in an actual open pathway in the physical integrity of containment.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. In this case, the licensee failed to identify the applicable TS-required actions for the degraded LPSW piping.
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: Based on the examples provided in section 6.9 of the Enforcement Policy, dated January 14, 2022, "Inaccurate and Incomplete Information or Failure to Make a Required Report," the performance deficiency was determined to be a SL IV violation. Specifically, example 6.9 states that a SL IV violation involves a licensee failure to make a report required by 10 CFR 50.73.
Violation: TS 5.4.1.a, Procedures, required, in part, that written procedures be established, implemented, and maintained covering activities related to procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978. Regulatory Guide 1.33, Section 1(h),
Administrative Procedures, required procedures addressing log entries, which was partially implemented by AD-OP-ALL-0112, Revision 0. AD-OP-ALL-0112, section 5.1.1.9 and 5.1.3.3, required, in part, that operators make log entries of entry and exit from TS LCOs.
Contrary to the above, the licensee failed to make operator narrative log entries for the entry and exit from TS LCO 3.0.3 on June 24 and 25, 2023.
10 CFR 50.73, "Licensee Event Report System," section (a)(2)(i)(B) requires, in part, that a licensee shall submit an LER for any operation or condition which was prohibited by the plants Technical Specifications within 60 days of the event or condition.
Contrary to the above, the licensee failed to notify the NRC within 60 days of a condition prohibited by Technical Specifications associated with a piping failure on the LPSW system in containment on June 24, 2023.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Report a Condition That Could Have Prevented Fulfillment of a Safety Function Associated with Online Reactor Building Cooling Unit Cleaning Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV 05000270/2023004-03 Applicable Open/Closed The inspectors identified a Severity Level IV NCV of 10 CFR 50.73(a)(2)(v) when the licensee failed to issue a licensee event report (LER) within 60 days of a condition that could have prevented fulfillment of a safety function. Specifically, on August 25, 2023, the licensee received an NRC comprehensive engineering team inspection report, which documented a condition that could have prevented fulfillment of a safety function associated with a procedure violation for the online cleaning of reactor building cooling units (RBCUs), but no NRC notifications were made.
Description:
On August 25, 2023, the NRC issued inspection report 05000269, 270, 287/2023011, Oconee Nuclear Station - Comprehensive Engineering Team Inspection (CETI) Report (ADAMS Accession No. ML23236A615). Within this report, Green NCV 05000270/2023011-01 was titled, Inappropriate Procedure, Instructions and Evaluation for Online RBCE Cleaning. The details of the documented violation established that Technical Specification 3.6.1, Containment, Condition A, applied during multiple periods of RBCU maintenance in April 2023. This condition was not recognized by Duke due to inappropriate procedural guidance governing the maintenance activity. Reference the inspection report for further details.
The inspectors noted that NUREG-1022, Revision 3, section 3.2.7, discusses reporting requirements for events or conditions that could have prevented fulfillment of a safety function. The applicable requirement for 60-day licensee event reports for this condition is 10 CFR 50.73(a)(2)(v). Accordingly, the guidance states that a report is required when:
1. there is a determination that the structure, system, or component (SSC) is inoperable in a required mode or other specified condition in the TS Applicability, 2. the inoperability is due to one or more personnel errors, including procedure violations; equipment failures; inadequate maintenance; or design, analysis, fabrication, equipment qualification, construction, or procedural deficiencies, and 3. no redundant equipment in the same system was operable.
The first two conditions are established in the CETI NCV write-up. The third condition is established in the Oconee UFSAR section 1.2.2.3, containment system, and various licensing bases references, such as Technical Specification Bases section 3.6.1, where the safety function of containment is discussed as being able to withstand the pressures and temperatures of the limiting accident without exceeding the leakage limits. It is clear from these references that the unit consists of a single containment and no redundant systems exist to accomplish that safety function if TS 3.6.1 applies due to the condition of containment being inoperable.
Inspectors also noted that, on July 20, 2023, during onsite discussions with the CETI team pertaining to the procedure and TS issues surrounding the RBCU cleaning activity, the licensee wrote NCR 2480047 with the subject, CETI23 Unit 2 Containment Inoperable for >1 hour During 2A RBCU Cleaning Restoration. Assignment 2 within this NCR was created on July 23, 2023, to review the reportability of the subject condition. The assignment was finally completed on February 5, 2024, and concluded that the condition was reportable under 10 CFR 50.73(a)(2)(v) for the reasons above. The licensee plans to submit an LER by the end of February 2024.
Corrective Actions: On January 31, 2024, the licensee wrote NCR 2503208, which acknowledged that a condition that could have prevented fulfillment of a safety function occurred during the RBCU cleaning activities and that an LER was required.
Corrective Action References: 2503208, 2480047
Performance Assessment:
The NRC determined that this violation was associated with a previously documented finding assessed using the significance determination process.
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: Based on the examples provided in section 6.9 of the Enforcement Policy, dated January 14, 2022, "Inaccurate and Incomplete Information or Failure to Make a Required Report," the performance deficiency was determined to be a SL IV violation. Specifically, example 6.9 states that a SL IV violation involves a failure to make a report to the NRC in accordance with 10 CFR 50.73.
Violation: 10 CFR 50.73, "Licensee Event Report System," section (a)(2)(v) requires, in part, that the licensee to submit a Licensee Event Report (LER) within 60-days, respectively, for any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:
(A) Shut down the reactor and maintain it in a safe shutdown condition;
(B) Remove residual heat;
(C) Control the release of radioactive material; or
(D) Mitigate the consequences of an accident.
Contrary to the above, the licensee failed to submit an LER within 60 days for a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Specifically, since at least August 25, 2023, Duke was aware of the NRC conclusion that TS 3.6.1 applied and containment was inoperable in some configurations during RBCU cleaning activities due to an inappropriate procedure. Therefore, an LER was required within 60 days.
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 October 17, 2023, the inspectors presented the construction of an ISFSI inspection results to Steven Snider, Site Vice President, and other members of the licensee staff.
- On November 9, 2023, the inspectors presented the Unit 2 ISI inspection results to Steven Snider, Site Vice President, and other members of the licensee staff.
- On November 16, 2023, the inspectors presented the RP occupational radiation safety inspection results to Steven Snider, Site Vice President, and other members of the licensee staff.
- On February 1, 2024, the inspectors presented the integrated inspection results to Steven Snider, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.04 Corrective Action 2457701, 2450253, 2477075, 2472049, 2495192
Documents
Drawings O-1478B Piping Layout Basement Floor Sections Reactor Building 49
O-1478D Piping Layout Basement Floor Partial Plan & Sections 49
Reactor Building
OFD-101A-2.2 Flow Diagram of High Pressure Injection System Storage 46
Section
OFD-102A-2.1 Flow Diagram of Low Pressure Injection System Borated 63
Water Supply and LPI Pump Suction
OFD-102A-2.2 Flow Diagram of Low Pressure Injection System LPI Pump 54
Discharge
OFD-102A-2.3 Flow Diagram of Low Pressure Injection System Core Flood 25
OFD-104A-1.2 Flow Diagram of Spent Fuel Cooling System Purification 25
Loop
OFD-121D-2.1 Flow Diagram of Emergency Feedwater System Rev. 42
OFD-122A-2.4 Flow Diagram of Main Steam Station (Emergency FDW Rev. 24
Pump Turbine Steam Supply & Exhaust)
OFD-127C-2.1 Flow Diagram of Nitrogen System (Nitrogen Supply to Close Rev. 8
MS-93 During AFIS Actuation)
Miscellaneous OSC-6104 Loss of Main Feedwater Event Mitigation Requirements Rev. 7
OSS-0254.00-00- (Mech) Design Basis Specification for the Emergency Rev. 059
1000 Feedwater System
OSS-0254.00-00- (Mech) Design Basis Spec for the Low Pressure Injection 069
28 and Core Flood System (LPI)
Procedures OP/2/A/1104/004 Low Pressure Injection System 177
OP/2/A/1106/006 Emergency Feedwater System Rev. 119
71111.05 Calculations OSC-10816 RONS TCCA A and B Area Basis for AD-EG-ALL-1520 005
OSC-9293 NFPA 805 Transition Radioactive Release G-1 Table 004
OSC-9314 NFPA 805 Transition Risk-Informed Performance-Based 006
Fire Risk Evaluation
OSC-9659 Oconee Nuclear Safety Capability Assessment for Units 1, 011
2, and 3
Inspection Type Designation Description or Title Revision or
Procedure Date
Corrective Action 2275773, 2276442, 2467200, 2464995, 2479435, 2426919,
Documents 2444366, 2479100
Fire Plans CSD-ONS-PFP-Pre-Fire Plan for U1 Auxiliary Building Elevation 796 001
CSD-ONS-PFP-Pre-Fire Plan for U2 Reactor Building All Elevations 000
2RB
CSD-ONS-PFP-Pre-Fire Plan for U2 Turbine Building Elevation 796 001
CSD-ONS-PFP-Pre-Fire Plan for U3 Auxiliary Building Elevation 783 001
O-0310-FZ-008 Auxiliary Building Unit 3 Fire Protection Plan Fire Area & 3
Fire Zone Boundaries Plan at EL 783+9
O-0310-FZ-009 Auxiliary Building Unit 1 Protection Plan Fire Area & Fire 3
Zone Boundaries Plan at EL 796+6 & 797+6
O-0310-FZ-029 Turbine Building - Units 2 Fire Protection Plan Fire Area & 2
Fire Zone Boundaries Plan at Mezzanine EL 796+6
O-0310-K-006 Auxiliary Building Unit 3 Fire Protection Plan & Fire Barrier, 13
Flood & Pressure Boundaries Plan at EL 783+9
O-0310-K-007 Auxiliary & Reactor Building Unit 1 Fire Protection Plan & 17
Fire Barrier, Flood & Pressure Boundaries Plan at EL 796+6
& 797+6
O-0310-L-005 Fire Protect Urbine BLDG Unit 2 EL 796+6 013
Miscellaneous AD-WC-ONS-ONS Shutdown Risk Management 4
20
OFD-100A-2.4 Flow Diagram of Reactor Coolant System R.C. Pump Motor 22
Drain System
OSS-0254.00-00- (Mech) Design Basis Specification for Fire Protection 47
4008
Procedures AD-FP-ALL-1520 Transient Combustible Control 0
AD-FP-ALL-1522 Duties of a Compensatory Fire Watch 1
CSD-ONS-FS-Standard Operating Guide Fires Located Within a 001
016 Contaminated RCA/RCZ
CSD-ONS-FS-Standard Operating Guide Fire Inside Containment 000
019
Inspection Type Designation Description or Title Revision or
Procedure Date
MP/0/A/1705/040 Fire Protection - Periodic Inspection of SLC Required Fire 002
71111.08P Engineering 180-9371235-000 Oconee 2R31 Baffle-to-Former, Flow Distributor, and Lower Rev. 1
Evaluations Core Barrel Bolt Ultrasonic Examinations
NDE Reports Steam Generator Degradation Assessment 0
03-9315755 Steam Generator Tube - Rolled Tube Plug & Stabilizer 8
Installation (ZR) Field Procedure
O-ISISG-Fifth Interval Steam Generator Inservice Inspection Plan 1
0169.030.0050
71111.11Q Corrective Action 2047523
Documents
Miscellaneous OFD-101A-2.4 Flow Diagram of High Pressure Injection System Charging 48
Section
OP-OC-16JT-06 PZR Cooldown JITT Lesson Plan 02
Procedures AD-TQ-ALL-0420 Conduct of Simulator Training and Evaluation 19
OP/1/A/1102/004 Operation At Power 163
OP/1/A/1103/005 Pressurizer Operation 066
OP/1/A/1106/002 FDWPT Operation 044
B
71111.12 Corrective Action 2490763, 2490862, 2478801, 2411493, 2403757, 2491077
Documents
Miscellaneous AD-EG-ALL-1210 Maintenance Rule Program Rev. 5
OSC-5771 PRA Risk Significant SSCs for Maintenance Rule Rev. 005
OSC-6551 PRA Analysis of Maintenance Rule Availability Performance Rev. 002
Criteria
OSS-0254.00-00- 100kV Alternate Power System Design Basis Document Rev. 020
2011
Procedures AP/1/A/1700/022 Loss of Instrument Air Rev. 031
Work Orders 20257082, 20257071, 20257113
71111.13 Miscellaneous AR 02490312 NRC Walkdown items identified for 3A LPI Pump Room
Procedures OP/3/A/1104/004 LPI System Fill and Startup 041
B
PT/3/A/0203/006 Low Pressure Injection Pump Test - Recirculation 094
A
Inspection Type Designation Description or Title Revision or
Procedure Date
Work Orders 20437227 01 3LPI-PU-0001 A LPIP Seal Housing Leak Off Line Boron
Deposit
71111.15 Calculations OSC-300 Containment Volume and Heat Sink in the Reactor Building 011
OSC-5986 Analysis of Miscellaneous Civil Components 048
OSC-9610 Evaluation of RBS, LPI, and HPI Systems for Generic Letter 002
08-01
Corrective Action 2459768, 2378192, 1787325, 2491695, 2493300, 2493304,
Documents 2492966, 2491967, 2463200, 2284431, 2481037, 2407794,
2302691, 2388222, 2312362, 2323011
Drawings OFD-101A-2.3 Flow Diagram of High Pressure Injection System Charging 34
Section
OFD-102A-1.2 Flow Diagram of Low Pressure Injection System LPI Pump 62
Discharge
OFD-102A-2.1 Flow Diagram of Low Pressure Injection System Borated 63
Water Supply and LPI Pump Suction
OFD-102A-2.2 Flow Diagram of Low Pressure Injection System LPI Pump 54
Discharge
P019-040-100 Oconee Water Shields 48 Model 2
Engineering 0000100905 Install manual remote operators on HPI suction block valves
Changes 2-HP-103, 2HP-107, and 2HP-993 for HELB and replace the
actuator for 2HP-107
Miscellaneous OSS-0254.00-00- (Mech) Design Basis Spec for the Low Pressure Injection 069
28 and Core Flood System (LPI)
OSS-0254.00-00- (Mech) Design Basis Specification for the Reactor Building 029
1034 Spray System
SD 1.3.9 Oconee-Specific Containment Materiel Control and Storage 22
Work Requests WR: 20258207, 20258208
Operability 2366042, 2495280
Evaluations
Procedures AD-EG-PWR-Boric Acid Corrosion Control Program - Implementation 6
1611
AD-MN-ALL-0006 Fluid Leak Management 5
AD-OP-ALL-0105 Operability Determinations 7
OP/1/A/1104/004 LPI System Fill and Startup 038
Inspection Type Designation Description or Title Revision or
Procedure Date
B
OP/2/A/1102/028 Reactor Building Tour and Containment Materiel Controls 021
OP/2/A/1104/004 Low Pressure Injection System 177
PT/1/A/0203/012 HPI/LPI/RBS Piping Venting 015
PT/2/A/0150/067 2LP-40 & 2LP-41 Leak Test 010
PT/2/A/1103/015 Reactivity Balance Procedure (Unit 2) 079
TE-MN-PWR-Inspection, Assessment, and Cleanup Of Boric Acid On 2
0006 Plant Materials
Work Orders 20547398, 20460059, 20535205, 20584157, 20513131,
20475586
01969469
71111.20 Corrective Action 2492143, 2492220, 1949528, 2492629
Documents
Drawings OFD-102A-2.1 Flow Diagram of Low Pressure Injection System Borated 63
Water Supply and LPI Pump Suction
OFD-102A-2.2 Flow Diagram of Low Pressure Injection System LPI Pump 54
Discharge
Miscellaneous Clearance PRT-2-23-PROT 2B LPI-0254
ONEI-0400-587 Oconee 2 Cycle 32 Final Core Load Map
ONEI-0400-591 Oconee 2 Cycle 32 Core Operating Limits Report 000
ONTC-0-101A-Unit 1&2 Spent Fuel Pool Level vs Temp Refueling Outage 027
0008-001 Specific Requirements for SSF RCMU Operability
OSC-11714 Oconee Nuclear Station Cycle Specific Decay Heat 015
Calculations
OSC-11847 Oconee Low Pressure Turbine Replacement Heavy Lift Risk 004
Evaluation
OSC-1594 NUREG-0612 - Control of Heavy Loads - Load Drop 002
Analysis
OSC-619 Analysis for Use of Spent Fuel Pool Inventory for SSF 057
OSS-0254.00-00- (Mech) Design Basis Specification for the Spent Fuel 034
1006 Cooling System
Procedures AD-WC-ONS-ONS Shutdown Risk Management 4
20
AP/1-Loss of SFP Cooling and/or Level 026
Inspection Type Designation Description or Title Revision or
Procedure Date
2/A/1700/035
IP/0/B/3002/002 A Polar Crane Inspection and Preventive Maintenance 015
MP/0/A/1150/002 Reactor Vessel - Closure Head - Removal 068
MP/0/A/1150/002 Reactor Vessel - Closure Head - Movement of Head From 012
G or To Stand
MP/0/A/1710/017 Crane - Whiting - Polar - Inspection 018
B
MP/0/A/3005/012 Containment Inspection/Close Out Procedure 020
MP/0/A/3007/090 Westinghouse Spent Fuel Pool Supplemental Cooling 001
MP/0/B/1710/022 Operation of Reactor Building Polar Crane and Auxiliary 039
Hoist (CRD Crane)
OP/0/A/1108/001 Spent Fuel Pool Level vs Temperature Curves 018
B
OP/1-SF Cooling System 113
2/A/1104/006
OP/1-SFP Makeup 024
2/A/1104/006 C
OP/2/A/1102/001 Controlling Procedure for Unit Startup 288
OP/2/A/1102/010 Controlling Procedure For Unit Shutdown 230
OP/2/A/1102/015 Filling And Draining FTC 084
OP/2/A/1102/028 Reactor Building Tour and Containment Materiel Controls 021
OP/2/A/1103/011 Draining And Nitrogen Purging RCS 013
OP/2/A/1104/004 Low Pressure Injection System 177
OP/2/A/1107/002 Normal Power 078
OP/2/A/1108/001 Curves and General Information 118
OP/2/A/1502/007 Operations Defueling/Refueling Responsibilities 084
PT/2/A/0600/001 Instrument Surveillance Prior to Mode Change 046
B
PT/2/A/0630/001 Mode Change Verification 033
Work Orders 20584215
71111.24 Corrective Action 2490367, 2459905, 2475902, 2482987, 2493373, 2493350,
Documents 2300234, 2472973, 2473883, 2494083
Drawings OFD-102A-2.1 Flow Diagram of Low Pressure Injection System Borated 63
Inspection Type Designation Description or Title Revision or
Procedure Date
Water Supply and LPI Pump Suction
OFD-116C-2.1 Flow Diagram of Reactor Building Hydrogen Purge System 024
OFD-116N-1.1 Flow Diagram of Standby Shutdown Facility HVAC 11
OM-254-0385.008 (DMV-1397) Nozzle Type Relief Valve A
OSC-4156 U2 LPI Heat Exchanger Performance Calculation 027
Miscellaneous EC 423349
AD-QC-ALL-0101 Quality Control Inspection Program for Modifications and 3
Maintenance Activities
OSS-0254.00-00- (Mech) Design Basis Specification for the Standby 027
1009 Shutdown Facility HVAC System
Procedures IP/0/A/2001/003/ Inspection and Maintenance Of Type HK Metal-Clad 029
C Switchgear, Associated Bus, And Disconnects
MP/0/A/1300/069 Maintenance Support for SSF Submersible Pump Test 2
OP/0/A/1107/011 Removal and Restoration of 4160V Main Feeder Buses 014
E
OP/0/A/1600/002 Standby Shutdown Facility Heating, Ventilation and Air 039
Conditioning System Operation (HVAC)
PT/0/A/0400/015 Submersible Pump Tests 26
PT/0/A/0400/016 SSF HVAC System Flow Test 022
PT/2/ A/0151/060 Penetration 60 Leak Rate Test 14
PT/2/A/0150/030 Containment Verification 003
PT/2/A/0151/061 Penetration 61 Leak Rate Test 13
PT/2/A/0251/069 LPI Cooler Test 011
PT/2/B/0200/005 Reactor Coolant Pump Motor Run 008
Work Orders 20629409, 20632933, 20460059, 20594448, 20583997,
20584043
71124.01 Procedures AD-RP-ALL-0002 Radiation and Contamination Surveys Rev. 3
Radiation ONS-M-Letdown Storage Tank 04/26/2022
Surveys 20220427-1
71124.03 Corrective Action AR 02495513 CRVS Carbon Filter Sample Surveillance Missed in 2019 11/21/2023
Documents (NRC Identified)
Resulting from
Inspection Type Designation Description or Title Revision or
Procedure Date
Inspection
Work Orders 20300709 01 Control Room Filter System Test 11/04/2019
71124.04 Corrective Action AR # 02425299, Various
Documents 02454531,
2467670,
2414597,
2442165,
2463709,
2464919,
2479854,
2480929, and
2488512
71124.05 Corrective Action ARs #02420981, Various
Documents 02434482,
2450912,
2459545,
2477133,
2479577,
2491397, and
2487525
Corrective Action AR #02491397 ONSI 305 Source Documentation Unavailable 10/23/2023
Documents AR #02494193 WO #20282613-02 Record not complete 11/09/2023
Resulting from
Inspection
Work Orders WO #20282613-1RIA-57& 58 Hi-Range RB Monitor Calibration
30