IR 05000269/2022004

Integrated Inspection Report 05000269 2022004 and 05000270/2022004 and 05000287/2022004

ML23030B323
Person / Time
Site:	Oconee
Issue date:	02/02/2023
From:	Eric Stamm Division Reactor Projects II
To:	Snider S Duke Energy Carolinas
References
IR 2022004
Download: ML23030B323 (26)

Text

February 1, 2023

SUBJECT:

OCONEE NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000269/2022004 AND 05000270/2022004 AND 05000287/2022004

Dear Steven Snider:

On December 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Oconee Nuclear Station. On February 1, 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.

Three findings of very low safety significance (Green) are documented in this report. These findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555- 0001; with copies to the Regional Administrator, Region 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. 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, Signed by Stamm, Eric on 02/01/23

Eric J. Stamm, Chief Reactor Projects Branch 1 Division of Reactor Projects

Docket Nos. 05000269 and 05000270 and 05000287 License Nos. DPR-38 and DPR-47 and DPR-55

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000269, 05000270 and 05000287

License Numbers: DPR-38, DPR-47 and DPR-55

Report Numbers: 05000269/2022004, 05000270/2022004 and 05000287/2022004

Enterprise Identifier: I-2022- 004- 0026

Licensee: Duke Energy Carolinas, LLC

Facility: Oconee Nuclear Station

Location: Seneca, South Carolina

Inspection Dates: October 1, 2022, to December 31, 2022

Inspectors: J. Nadel, Senior Resident Inspector A. Ruh, Resident Inspector N. Smalley, Resident Inspector P. Cooper, Senior Reactor Inspector S. Downey, Senior Reactor Inspector J. Viera, Senior Operations Engineer

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 Use Procedure Appropriate to the Circumstance for Low Pressure Injection System Alignment Results in Excessive Reactor Coolant System Leakage Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71111.04 NCV 05000269/2022004-01 Complacency Open/Closed A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50,

Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to use a procedure appropriate to the circumstances for an activity affecting quality. Specifically, the procedure for aligning the B train of low pressure injection (LPI) to the emergency safeguards (ES) alignment resulted in a loss of primary inventory that lowered pressurizer level by six inches. This loss of inventory event resulted in entry into an abnormal procedure for excessive reactor coolant system (RCS) leakage.

Failure to Test Emergency Condenser Circulating Water First Siphon Under Suitable Environmental Conditions Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.15 Systems NCV 05000269,05000270,05000287/2022004- 02 Open/Closed The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when the licensee failed to perform testing, under suitable environmental conditions, to demonstrate that the emergency condenser circulating water (ECCW) first siphon would perform satisfactorily in service for the entire range of approved operating lake level conditions.

Failure to Implement Locked High Radiation Area Controls Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.12] - Avoid 71152A Radiation Safety NCV 05000287/2022004-03 Complacency Open/Closed The inspectors identified a Green finding and associated non-cited violation (NCV) of Technical Specification 5.4.1(a), Procedures, when the licensee failed to implement controls for a locked high radiation area (LHRA) as required by procedure AD-RP-ALL-0004,

"Radiological Posting and Labeling." Specifically, on September 27, 2022, NRC inspectors found the access door to auxiliary building room 254, Unit 3 spent fuel demineralizer/filter room, unlocked and unguarded, and personnel access was uncontrolled with the area exhibiting accessible dose rates of at least 1.4 rem/hr at 30 cm from the radiation source.

Additional Tracking Items

None.

PLANT STATUS

Unit 1 began the inspection period at or near 100 percent rated thermal power (RTP). On October 16, 2022, the unit began end of cycle power reductions for a scheduled refueling outage and was shut down for the outage on October 28, 2022. The unit reached 100 percent RTP on November 26, 2022, and remained at or near 100 percent RTP for the remainder of the inspection period.

Unit 2 operated at or near 100 percent RTP for the entire inspection period.

Unit 3 operated at or near 100 percent RTP until November 6, 2022, when power was reduced to 87 percent RTP for repairs to a heater drain pump. Power was returned to 100 percent RTP on November 8, 2022, and remained there for the rest of the inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (3 Samples)

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

(1) Unit 1 power path alignment during reduced inventory operations on November 17, 2022

(2) Unit 1 high pressure injection (HPI) system following restoration from refueling outage on November 21, 2022

(3) Essential siphon vacuum system during 2A train corrective maintenance on December 15, 2022

Complete Walkdown Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated system configurations during a complete walkdown of the Unit 1 low pressure injection system during the refueling outage between October 28 and 31, 2022, with additional in-office reviews.

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 27, 2022

(2) Fire zone 33: Unit 2 4160V switchgear room on October 27, 2022

(3) Fire zone 122: Unit 1 containment on November 2, 2022

(4) Fire zone 77: Unit 3 auxiliary building 200 level hallway on December 21, 2022

(5) Fire zone 103: Unit 2 east penetration room on December 21, 2022

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) Unit 1 A and B low pressure injection coolers

===71111.08P - Inservice Inspection Activities (PWR)

PWR Inservice Inspection Activities Sample (IP Section 03.01)===

(1) The inspectors evaluated pressurized-water reactor (PWR) non-destructive testing by reviewing the following examinations from October 31 to November 10:

Eddy Current Examination a.

Steam Generator (SG) 1A, Tube R40C105 and R140C50, American Society of Mechanical Engineers (ASME) Class 1 (observed)b.

SG 1B, Tube R72C2 and R75C2, ASME Class 1 (observed)

Liquid Penetrant Examination a.

1-51A-0135-113, pipe to elbow weld, ASME Class 1. This included a review of associated welding activities.

b.

1-51A-0135-114, elbow to valve weld, ASME Class 1. This included a review of associated welding activities.

c.

Reactor Pressure Vessel Head Forging, ASME Class 1

Ultrasonic Examination a.

Weld 1PDB1-11, nozzle to safe end weld, ASME Class 1 b.

Weld 1RC-201-121, pipe to safe end weld, ASME Class 1 c.

Weld 1PDA1-11, nozzle to safe end weld, ASME Class 1 d.

1-PDA1-47, high pressure injection nozzle safe end, ASME Class 1 e.

Base metal in 1A1 high pressure injection line piping from weld 1RC-199-154 to weld 1RC-199-149, ASME Class 1

The inspectors also evaluated the licensees boric acid control program performance.

71111.11A - Licensed Operator Requalification

Program and Licensed Operator Performance

Requalification Examination Results (IP Section 03.03) (1 Sample)

(1) The licensee completed the annual requalification operating 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." During the week of November 14, 2022, the inspector performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with IP 71111.11, "Licensed Operator Requalification Program." These results were compared to the thresholds established in Section 3.03, "Requalification Examination Results," of IP 71111.11.

The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam completed on April 4, 2022.

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 reactor shutdown on October 28, 2022.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

(1) The inspectors observed and evaluated just in time simulator training for an operating crew using simulator exercise guide OP-OC-15JT-05 on October 26, 2022.

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) Nuclear condition reports (NCRs) 2448516, 2448570, Unit 1 turbine building to auxiliary building submarine flood door and surrounding concrete structure leaking

(2) NCR 2445019, significant high pressure service water piping leak during excavation work

Quality Control (IP Section 03.02) (1 Sample)

The inspectors evaluated the effectiveness of maintenance and quality control (QC)activities to ensure the following SSC remains capable of performing its intended function:

(1) QC verification of the installation of the standby shutdown facility (SSF) emergency diesel generator (EDG) lube oil immersion heater starter in accordance with work order (WO) 20565773 Aging Management (IP Section 03.03) (1 Sample)

The inspectors evaluated the effectiveness of the aging management program for the following SSCs that did not meet their inspection or test acceptance criteria:

(1) NCR 2447967, 1B2 high pressure injection nozzle pipe to safe end weld with unacceptable flaw on November 1, 2022

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (1 Sample)

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

(1) Unit 1 yellow shutdown risk during lowered inventory operations, on November 1, 2022

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1) NCR 2448656, CT-1 non-segregated 4160V startup bus bolted connections need to be replaced

(2) NCR 2448967, functionality evaluation of the upper thermal shield restraint due to broken bolt and retaining clip identified in O1R32 outage

(3) NCR 2449194, boron residue found on 1B2 reactor coolant pump bearing thermocouple

(4) NCR 2449176, corroded low pressure service water piping in Unit 1 containment

(5) NCR 2450052, condenser circulating water pump flange seal test PT/X/A/0261/021 test method inadequate

(6) NCR 2449282, motor operated valve 1LPSW-139 set-up below acceptance criteria

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)

The inspectors evaluated the following temporary or permanent modification:

(1) NCR 2448656, CT-1 non-segregated 4160V startup bus bolted connections changed from 1/2-inch to 9/16-inch Belleville washers

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (4 Samples)

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

(1) IP/0/A/3011/015, Removal and Replacement of Motor Control Center, Panelboards, and Remote Starter Components, following replacement of the SSF EDG lube oil/jacket water immersion heater starter, WO 20565773

(2) IP/1/A/0275/012 A, 1FDW -44 Calibration Check, following replacement of Moore I/P converter with ControlAir, WO 20542872

(3) OP/A/1600/010, Operation of the SSF Diesel Generator, following maintenance on the SSF diesel generator lube oil pressure switches, on October 25, 2022

(4) Replace 1B2 reactor coolant pump radial bearing resistance temperature device (RTD) with new thermocouple RTD, WO 20568663

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated refueling outage U1R32 activities from October 28, 2022, to November 24, 2022.

71111.22 - Surveillance Testing

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

Inservice Testing (IP Section 03.01) (2 Samples)

(1) PT/1/A/0152/012, Low Pressure Injection System Valve Stroke Test, on November 7, 2022

(2) PT/1/A/0251/024, HPI Full Flow Test, on November 17, 2022

Containment Isolation Valve Testing (IP Section 03.01) (1 Sample)

(1) PT/1/A/0151/056, Penetration 56 Leak Rate Test, on November 19, 2022

FLEX Testing (IP Section 03.02) (1 Sample)

(1) FLEX submersible pump testing under WO 20547704, on October 26, 2022

71114.06 - Drill Evaluation

Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)

(1) Training drill 2022- 04 on October 18, 2022, which included operations shift D, emergency response organization teams 2 and 5, with participation from the emergency operations facility

OTHER ACTIVITIES - BASELINE

===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) Unit 3 spent fuel demineralizer and filter room locked high radiation area door left unsecured

===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 operations and maintenance department performance that might be indicative of a more significant safety issue.

INSPECTION RESULTS

Failure to Use Procedure Appropriate to the Circumstance for Low Pressure Injection System Alignment Results in Excessive Reactor Coolant System Leakage Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71111.04 NCV 05000269/2022004-01 Complacency Open/Closed A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to use a procedure appropriate to the circumstances for an activity affecting quality. Specifically, the procedure for aligning the B train of low pressure injection (LPI) to the emergency safeguards (ES) alignment resulted in a loss of primary inventory that lowered pressurizer level by 6 inches. This loss of inventory event resulted in entry into an abnormal procedure for excessive reactor coolant system (RCS) leakage.

Description:

On November 20, 2022, Unit 1 was at the end of a refueling outage in mode 5 with the LPI system in service for decay heat removal. The licensee attempted to realign the system to its ES alignment in preparation for transitioning to mode 4. Prior to the outage, the procedure for aligning LPI had been modified as a time-saving measure. Specifically, it left one train of LPI in the high pressure mode of decay heat removal while simultaneously aligning the opposite train for ES prior to starting any reactor coolant pumps. Step 2.10.2 of enclosure 4.7, ES Alignment, in procedure OP/1/A/1104/004, "Low Pressure Injection System," Revision 161, directed operators to open the B train LPI pump suction isolation valve, 1LP-8. When this valve was opened, it created a flow path between the high-pressure discharge side of the 1A LPI pump to the discharge side of the non-running 1B LPI pump.

Due to a single discharge check valve in that line that was not leak tight, this flow path pressurized the suction side of the 1B pump and lifted a suction relief valve. Soon after the step was completed, operators noticed pressurizer level decreasing while high activity waste tank level was increasing. Operators entered abnormal procedure OP/1/A/1700/002, "Excessive RCS Leakage," and took the required actions. After approximately 4.5 minutes, operators closed suction isolation valve 1LP-8, which stopped the leakage. Pressurizer level dropped approximately six inches during the event with a calculated leak rate of approximately 40 gallons per minute.

Residents reviewed a similar event that happened on Unit 2 in 2021 as documented in NCR 2407794. The 2021 event, which also resulted in entry into the excessive RCS leakage abnormal procedure, showed that the LPI pump discharge check valves were not an adequate boundary for similar alignments under similar conditions. The residents also identified additional valve alignments in enclosure 4.7 which connected the running and non-running LPI trains and for which a sound procedural basis could not be determined. These alignments likely contributed to the event.

Corrective Actions: OP/1/A/1104/004 will be revised to prevent similar events when aligning the LPI system during outages.

Corrective Action References: 2450253

Performance Assessment:

Performance Deficiency: The licensees failure to use a procedure appropriate to the circumstances while aligning the LPI system on November 20, 2022, was 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 Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the inappropriate procedure resulted in excessive RCS leakage that lowered pressurizer level by six inches while the unit was in mode 5.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Safety SDP. The performance deficiency involved shutdown operations and did not involve fuel handling errors. Using IMC 0609 Appendix G, 1, "Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings," Exhibit 2, Initiating Events Screening Questions, this issue screened to a Phase 2 evaluation because it was a loss of inventory event with leakage rates such that if the leakage were undetected and/or unmitigated it would cause the currently operating decay heat removal method to fail in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less. A regional Senior Reactor Analyst (SRA) performed an IMC 0609 Appendix G risk screening using 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, Worksheet 5, SDP for a PWR Plant - Loss of Inventory in [Plant Operating State]

POS I (RCS Closed).

The RCS was closed with level in the pressurizer. The LPI system was in service in the decay heat removal (DHR) alignment. Steam generators were intact, filled, and available for decay heat removal if needed. The plant was late in the outage (TW-Late),so decay heat levels were reduced. The SRA evaluated the sequences in worksheet 5 using IMC 0609 Appendix G, Attachment 2, Table 3 - Initiating Event Likelihood (IELs) for Loss of Inventory Precursors. Note, a shutdown loss of inventory event is considered an event where the loss of inventory results in the loss of the in -service method of decay heat removal. IEL was set to 4 because operators would have approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to isolate the leak path, leak isolation can be performed in less than half the time available, LPI pump indication and reactor water level instruments were functional and available, and DHR could be restored.

Using IMC 0609 Appendix G, Attachment 2, Table 9 - Counting Rule Worksheet, the issue screened to very low safety significance (Green).

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. In this case, when implementing the procedure change, the organization was complacent about the reliance on check valves in the LPI system as a pressure/leakage boundary and failed to recognize the inherent risk of this alignment despite past examples of site-specific check valve leakage in this system.

Enforcement:

Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed 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. This requirement is implemented, in part, by Oconee procedure OP/1/A/1104/004, "Low Pressure Injection System," Revision 161. Contrary to the above, on November 20, 2022, the LPI system was aligned in accordance with OP/1/A/1104/004, Revision 161, and this procedure was not appropriate to the circumstances for the alignment activity, which is an activity affecting quality. Specifically, the procedure resulted in excessive RCS leakage that lowered pressurizer level by six inches while the unit was in mode 5.

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

Failure to Test Emergency Condenser Circulating Water First Siphon Under Suitable Environmental Conditions Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.15 Systems NCV 05000269,05000270,05000287/20220 04- 02 Open/Closed The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when the licensee failed to perform testing, under suitable environmental conditions, to demonstrate that the emergency condenser circulating water (ECCW) first siphon would perform satisfactorily in service for the entire range of approved operating lake level conditions.

Description:

The ECCW system uses a siphon (also called first siphon) to ensure water can be maintained from Lake Keowee to the low pressure service water (LPSW) pumps following a loss of condenser circulating water (CCW) pump operation during a loss -of-offsite power condition. In order to maintain waterflow via siphon action, proper operation of the essential siphon and vacuum system is routinely tested to ensure its air removal rate exceeds the expected rate of air in-leakage entering the ECCW piping. Typically, routine ECCW system tests are conducted at lake levels that are 796 to 800 feet above mean sea level. At these lake levels, an in-leakage pathway through a mechanical joint in each of the CCW pumps column is typically submerged which prevents air in-leakage through the joint during the test. When the lake level is below 792.7 feet (such as during drought conditions when lake level could be at a minimum of 790 feet), the joint can become uncovered and can become an additional source of air in-leakage that is not evident during routine ECCW tests. Testing per PT/1,2,3/A/0261/021, CCW Pump Flange Seal Test, is conducted to validate the integrity of the flange seals. The test sought to internally pressurize the volume between the 96-inch discharge butterfly valve and static water level inside the CCW pump column with air and examine the exterior of the column for bubbles indicating leakage past the rubber seal.

Proper operation of the seal was important to safety because current site calculations and operational allowances presumed the ECCW first siphon supported the operability of the LPSW system for lake levels down to 790 feet.

In November 2022, inspectors identified that the test methodology in PT/1,2,3/A/0261/021 did not establish suitable environmental conditions for the purpose of the test.

The test procedure instructed technicians to connect an air compressor to the bearing water lubrication line of each pump and to pressurize the line to approximately 3 psig. Inspectors noted that the bearing water lubrication line was not directly open to the pump column space that the test was attempting to pressurize. Instead, the line was connected to a volume defined by the shaft packing and a series of cover pipes that surrounded the pump shaft. To pressurize the pump column from the lubrication line, air would need to leak through small clearances at cover pipe connections at the first bearing housing or stuffing box, and the 96-inch discharge butterfly valve would need to maintain sufficient leak tightness. However, these elements were not verifiable using the existing test procedure and instrumentation setup since only the pressure within the cover pipe could be monitored. Essentially, there was no direct way to measure the actual pressure in the pump column to ensure the interior of the flange seal had been uncovered when examining the exterior for bubbles. Oconees quality assurance program, described in DUKE-QAPD-001-A, Section D17.3.2.8, Test Control, required that the conditions necessary to perform the specified testing must be considered in the preparation and review of procedures. The licensee entered the testing issue into the corrective action program and performed an operability determination. The operability determination assumed the flange seal was not airtight and concluded the ECCW system was operable, with margin, if the lake was maintained greater than 795 feet. At the time of the determination the lake level was greater than 797 feet and had been greater than 796 feet for at least the past three years.

Corrective Actions: The licensee coordinated with authorities to ensure Keowee lake levels remained greater than 795.1 feet, implemented a standing order to monitor lake level, and initiated action to develop a system design change to facilitate future testing and verification of pump flange seal integrity.

Corrective Action References: 2450052

Performance Assessment:

Performance Deficiency: The failure to prepare test procedure PT/1,2,3/0261/021 with conditions necessary to perform CCW pump flange seal testing, as required by DUKE-QAPD-001-A, was a performance deficiency. Specifically, the purpose of the test was to check CCW pump column joints for leakage, but the test failed to ensure the intended locations were pressurized due to the test method.

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. The failure to test the CCW pump column flanges for leakage, under suitable environmental conditions, adversely affected the objective of ensuring the reliability and capability of the ECCW first siphon to operate as designed. Specifically, the test was not capable of detecting a failed seal which could undermine design assumptions for the LPSW system within the range of approved operating lake level conditions.

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 Screening Questions, inspectors concluded the finding was Green based on a review of recent Keowee lake level data indicating the ECCW system would have maintained its operability even under the conservative assumption that the flange seals were completely ineffective.

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: 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed. Also, that test procedures shall include provisions for assuring that the test is performed under suitable environmental conditions. Contrary to the above, the licensee failed to perform testing, under suitable environmental conditions, to demonstrate that the ECCW first siphon would perform satisfactorily in service at lower than typical lake levels. Specific ally, the purpose of PT/1,2,3/A/0261/021 was to check CCW pump column joints for leakage, but the test failed to ensure the intended locations were pressurized due to the test method.

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

Failure to Implement Locked High Radiation Area Controls Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.12] - Avoid 71152A Radiation Safety NCV 05000287/2022004 - 03 Complacency Open/Closed The inspectors identified a Green finding and associated non-cited violation (NCV) of Technical Specification 5.4.1(a), Procedures, when the licensee failed to implement controls for a locked high radiation area (LHRA) as required by procedure AD - RP-ALL- 0004, "Radiological Posting and Labeling." Specifically, on September 27, 2022, NRC inspectors found the access door to auxiliary building room 254, Unit 3 spent fuel demineralizer/filter room, unlocked and unguarded, and personnel access was uncontrolled with the area exhibiting accessible dose rates of at least 1.4 rem/hr at 30 cm from the radiation source.

Description:

On September 27, 2022, during a tour of the auxiliary building, the inspectors identified that the caged door frame to the Unit 3 spent fuel demineralizer/filter room was broken, and that the licensee had provided a separate padlock and chain to secure the area.

Although the area was posted as a LHRA, the ins pectors examined the padlock and chain and determined that it was in an ineffective configuration. Specifically, the padlock secured one end of the chain to the caged door while leaving the other end of the chain wrapped around a nearby metal support witho ut a locking mechanism, rendering the area unlocked.

The area had been entered earlier that month on September 9, 2022, for a valve manipulation. At that time, the room was determined to exhibit accessible general area radiation dose rates of up to 1.4 rem /hr per radiation survey ONS -M-20220709-1, meeting the definition of a LHRA. Licensee procedure AD-RP-ALL- 0004, "Radiological Posting and Labeling," states that a barrier with a locked door is required for the boundary of a locked high radiation area except while the area is being accessed. After exiting the area, the technician re-installed the access control chain incorrectly. The inadequate access control configuration was not identified at that time by the peer check, nor was it identified during two subsequent weekly LHRA reviews by the licensee. Furthermore, no other access control methods (e.g., a door guard) were in place. The inspectors determined that the caged door frame had been broken for an extended period of time (more than one year) and that the padlock and chain were being used until the door frame could be repaired via the work request process. The inspectors interviewed Radiation Protection management, reviewed dose rate alarm records, and determined that it was unlikely that any unauthorized access occurred during the time the degraded barrier existed.

Corrective Actions: Upon notification by the inspectors, the licensee immediately reconfigured the padlock and chain to lock access to the area under supervisory oversight and entered this issue into its corrective action program (CAP). An extent of condition determination and verification of all other LHRA access controls was completed. The licensee also conducted a Prompt Investigation Response Team investigation as well as Human Performance and Organization and Programmatic Checklist evaluations. The original work request for the broken door frame was rescreened as CAP, since it had initially been screened non-CAP, and the door was repaired.

Corrective Action References: 2443366

Performance Assessment:

Performance Deficiency: The licensees failure to implement LHRA controls as required by procedure AD-RP-ALL-0004, "Radiological Posting and Labeling," 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, it resulted in the loss of exposure control to an area exhibiting elevated radiation dose rates.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP.

The finding was determined to be of very low safety significance (Green) because it was not related to ALARA [as low as reasonably achievable], did not result in an overexposure, did not present a substantial potential for overexposure, and the licensees ability to assess dose was not compromised.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. In this case, when initially implementing and subsequently reviewing the locked high radiation areas controls for auxiliary building room 254, the organization was complacent about the reliance on the presence of a padlock and chain that seemed to secure the area and failed to recognize that there were multiple ways for the padlock and chain to be improperly used to secure the area.

Enforcement:

Violation: Technical Specification 5.4.1(a), Procedures, states that written procedures shall be established, implemented, and maintained covering the activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 7, Procedures for Control of Radioactivity, includes access control to radiation areas. Licensee procedure AD-RP-ALL-0004, "Radiological Posting and Labeling,"

states, in part, that a barrier with a locked door is required for the boundary of a locked high radiation area except while the area is being accessed. Contrary to the above, from at least September 9, 2022, to September 27, 2022, access to auxiliary building room 254, Unit 3 spent fuel demineralizer/filter room, was not controlled per AD-RP-ALL-0004.

Specifically, on September 27, 2022, NRC inspectors identified that room 254 did not have a barrier with a locked door while the room was not being accessed by personnel and dose rates within the room were determined to meet LHRA conditions.

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 February 1, 2023, the inspectors presented the integrated inspection results to Steven Snider and other members of the licensee staff.

• On November 10, 2022, the inspectors presented the Unit 1 in-service inspection results to Steven Snider and other members of the licensee staff.

THIRD PARTY REVIEWS

Inspectors reviewed Institute on Nuclear Power Operations reports that were issued during the inspection period.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Corrective Action 2164850, O -08 -01643, 2396702, 2366042, 2373308,

Documents 2449312, 2451863, 2450272, 2438431, 2450253, 2451877,

26303

Drawings O -0702 One Line Diagram 6900V & 4160V STA Auxiliary Sys 40

O -0702 -A One Line Diagram 6900V & 4160V Auxiliary Sys 39

O -0703 -G One Line Diagram Station Auxiliary Circuits 600/208V L/C 97

1X8, 1X9, 1X10 & MCC 1XS1, 1XS2, 1XS3

O -703 -E One Line Diagram Station Auxiliary Circuits 600V/208V L/C 75

1X6 & MCC 1XI, 1XN, 1XP & 1XQ

OFD -101A -1.1 Flow Diagram of High Pressure Injection System Letdown 53

Section

OFD -101A -1.2 Flow Diagram of High Pressure Injection System Storage 48

Section

OFD -101A -1.3 Flow Diagram of High Pressure Injection System Charging 37

Section

OFD -101A -1.4 Flow Diagram of High Pressure Injection System Charging 51

Section

OFD -101A -1.5 Flow Diagram of High Pressure Injection System SSF 29

Portion

OFD -130A -2.1 Flow Diagram of Essential Siphon Vacuum System 12

Miscellaneous Clearance PSC -1 -22 -HP -1HP490 491 -0156

Clearance PRT -1 -22 -CT -1 2ND PWR -0339

Clearance PRT -1 -22 -CT - 5 -U1 MFB -0235(3)

Clearance PRT -1 -22 -REDUCED INV -0332

Clearance PRT -1 -22 -1LT -5 -0312

Clearance PRT -1 -22 -1B LPI TRAIN -0311

Clearance PRT -1 -22 -ADM DROP LPS -0315

Procedures OP/1/A/1104/002 HPI System 182

Work Orders 20513544, 20518531, 20285643, 20406681, 20237935

71111.05 Calculations OSC -9314 NFPA 805 Transition Risk -Informed Performance -Based 006

Fire Risk Evaluation

Corrective Action 2274272, 2407000

Documents

Inspection Type Designation Description or Title Revision or

Procedure Date

Fire Plans CSD-ONS-FS-Standard Operating Guide Fire Inside Containment 000

019

CSD-ONS-PFP-Pre-Fire Plan for U1 Auxiliary Building Elevation 796 001

1AB-0796

CSD-ONS-PFP-Pre-Fire Plan for U1 Reactor Building 000

1RB

CSD-ONS-PFP-Pre-Fire Plan for U2 Auxiliary Building Elevation 809 0

2AB-0809

CSD-ONS-PFP-Pre-Fire Plan for U2 Turbine Building Elevation 796 000

2TB-0796

CSD-ONS-PFP-Pre-Fire Plan for U3 Auxiliary Building Elevation 783 1

3AB-0783

O-0310-FZ-009 Auxiliary Building Unit 1 Fire Protection Plan Fire Area & Fire 3

Zone Boundaries Plan at EL 796+6 & EL 797+6

O-0310-FZ-029 Turbine Building Unit 2 Fire Protection Plan Fire Area & Fire 2

Zone Boundaries Plan at Mezzanine EL 796+6

O-0310-K-007 Fire Protect Aux Bldg Unit 1 EL 796+6 16

O-0310-L-005 Turbine Building Unit 2 Fire Protection Plan & Fire Barrier, 13

Flood, & Pressure Boundaries Plan at Mezzanine EL 796+6

O-310-K-06 Auxiliary Building - Unit 3 Fire Protection Plan & Fire Barrier, 12

Flood, & Pressure Boundaries Plan at EL 783+9

O-310-K-11 Auxiliary Building - Unit 2 Fire Protection Plan & Fire Barrier, 16

Flood, & Pressure Boundaries Plan at EL 809+3

Miscellaneous AD-FP-ALL-1520 Transient Combustible Control 0

AD-WC-ONS-ONS Shutdown Risk Management 3

20

OFD-124C-1.2 Flow Diagram of High Pressure Service Water System 40

Turbine Building

Work Orders 20239781, 20239784

71111.07A Calculations OSC-2693 Decay Heat Removal Cooler Performance Calculation 1

Methodology Verification

OSC-3993 U1 LI Heat Exchanger Performance Calculation 24

Drawings OM 201-0286.001 #46-128:3940 Decay Heat Coolers 17

Inspection Type Designation Description or Title Revision or

Procedure Date

Procedures IP/1/A/0203/001 C Unit 1 Low Pressure Injection System Accessible Flow 5

Instruments Calibration

PT/1/A/0251/069 LPI Cooler Test 12

71111.08P Corrective Action NCRs (by number) 02436679, 02447643, 02447680, 02447967

Documents

NDE Reports PT-22-001 Liquid Penetrant Examination of RPV Head Forging 10/31/2022

Surface

UT-22-017 Ultrasonic Examination of 1RC-201-121 (PDI-UT-3 for flaw 11/01/2022

disposition)

UT-22-024 Ultrasonic Examination of 1RC-201-121 10/31/2022

71111.11Q Procedures OP-OC-15JT-05 Night Shift Shutdown JITT 12a

OP/1/A/1102/010 Controlling Procedure for Unit Shutdown 236

71111.12 Calculations OSC-7966 EWST Inventory for Turbine Building Flood 010

OSC-8729 Supporting Vendor Analysis for Alloy 600 Mitigation - D2

Oconee Nuclear Station Units 1, 2 & 3 - HPI Evaluations

Supporting Environmental Fatigue Effects

Corrective Action 2445019, 2326989, 2336744, 2342922, 2431932

Documents

Drawings OFD -124C -1.4 Flow Diagram of High Pressure Service Water System 42

West Yard

OFD -124C -1.5 Flow Diagram of High Pressure Service Water System East 21

Yard and Visitors Center

OM 201 -597.001 Assembly & Detail for 2 -1/2" Pressure Injection Nozzle D2

Engineering 111894 QA -1 Weld Build Up on 1B2 Nozzle Safe End 4

Changes 421842 QA-1 Weld Overlay on 1B2 HPI Nozzle Safe End to Pipe 1

Weld

Miscellaneous AD -EG -ALL -1210 Maintenance Rule Program 4

Structural Integrity Oconee Unit 1 HPI Design Reconciliation & Qualification 0

Associates, Inc.

Report

201209.310

Structural Integrity Interim Assessment of Crack Growth - HPI Nozzle 1B2 1

Associates, Inc. Weld Overlay ONS1

Report

201209.401

Inspection Type Designation Description or Title Revision or

Procedure Date

Structural Integrity ONS1 1B2 HPI Nozzle - Safe End to Pipe Weld Overlay 0

Associates, Inc. Repair Compliance Matrix for ASME Code,Section XI,

Report Nonmandatory Appendix Q, 2017 Edition

201209.402

TE-MN-ALL-0510 Excavation and Backfill 1

Procedures MP/0/A/1705/008 A Fire Protection - Hydrant - Disassembly, Repair, and 012

Reassembly

OP/0/A/1104/011 High Pressure Service Water 108

OP/1/A/6101/009 Alarm Response Guide 1SA-09 052

Work Orders 20513217, 20418874, 20517775

71111.15 Calculations OSC -1306.06 Piping Analysis for Low Pressure Service Water System to D13

and from RBCU Cooling Coils

OSC -4553 Maximum Differential Pressures for LPSW Valves 7

OSC -4671 Units 1&2 LPSW System Benchmark 013

OSC -5665 Reactor Building Cooling Unit Performance Test Unit 1 071

OSC -5760 Generic Letter 89 -10 MOV Calculation for Unit 1 Butterfly 15

Valves at Oconee

Corrective Action 2449176, 2386632

Documents

Drawings OFD -124B -1.2 Flow Diagram of Low Pressure Service Water System 37

Reactor Building Cooling Units

OM 202 -0004.001 80X96 Type SAFV Pump 1

ONTC -1 -LPSW - 139 Test Acceptance Criteria for Valve 1LPSW -139 9

Miscellaneous ONEI -0400 -400 Revised RBCU Requirements in Support of 1-Year Mission 1

Times

Work Orders 20568729, 20398370, 20566826, 20566827

71111.19 Corrective Action 2430827, 2426484

Documents

Drawings O -422M -32 Instrument Detail Startup FDW Valve Control 1FDW -35 19

& 1FDW -44

OFD -121B -1.3 Flow Diagram of Feedwater System (Final Feedwater) 42

Engineering 421202

Changes

Inspection Type Designation Description or Title Revision or

Procedure Date

Procedures IP/1/A/0275/012 A Unit 1 Feedwater Control Valve Demand and Interlock 004

Calibration

Work Orders 20542872

71111.20 Corrective Action 1761227, 1794841, 2448302, 2450428, 2450570

Documents

Drawings O-422BB-6 Instrument Detail Reactor Vessel Level Loop A (1RC 13

LT0005A)

O-422BB-6.01 Instrument Detail Reactor Vessel Level Loop B (1RC 2

LT0005B)

OFD-104A-1.1 Flow Diagram of Spent Fuel Cooling System 68

OFD-104A-1.2 Flow Diagram of Spent Fuel Cooling System Purification 25

Loop

Engineering 421670 Allowable Alternate Power Configuration for a SFP Cooling 0

Changes Pump

Miscellaneous Clearance OPS-1-22-EL-1XL REMOVAL-0839

Clearance OPS-1-22-EL-REMOVAL MFB1-0849

Clearance OPS-1-22-EL-1TD REMOVAL-0841

Clearance OPS-1-22-EL-1X91XS2/1XS5-1655

Oconee 1 Cycle 33 Reload Core Design 10 CFR 50.59

Screen

ONEI-0400-301 Communication of Requirements for Oconee Technical 1

Specification 3.7.13 - Definitions, Boundary Conditions, and

Spent Fuel Pool Storage Qualifications

ONEI-0400-575 Oconee 1 Cycle 33 Final Core Load Map 0

ONEI-0400-583 Oconee 1 Cycle 33 Core Operating Limits Report 1

ONEI-0400-584 O1C33 Physics Test Manual 0

Procedures AD-MN-ALL-0002 Foreign Material Exclusion 15

AD-SY-ALL-0460 Managing Fatigue and Work Hour Limits 7

AD-WC-ALL-0420 Shutdown Risk Management 7

AD-WC-ONS-ONS Shutdown Risk Management 3

20

AP/1-Loss of SFP Cooling and/or Level 26

2/A/1700/035

AP/1/A/1700/026 Loss of Decay Heat Removal 30

Inspection Type Designation Description or Title Revision or

Procedure Date

IP/0/B/0200/027 A Reactor Level Instrumentation and Calibration 33

MP/0/A/1500/009 Defueling/Refueling Procedure 078

OP/0/A/1104/006 F SF Reverse Osmosis System Operation 010

OP/1-2/A/1104/006 SF Cooling System 113

OP/1-2/A/1104/009 RCW System 31

OP/1/A/1102/001 Controlling Procedure for Unit Startup 325

OP/1/A/1102/010 Controlling Procedure for Unit Shutdown 236

OP/1/A/1102/028 Reactor Building Tour and Containment Material Controls 017

OP/1/A/1104/004 Low Pressure Injection System 161

OP/1/A/1105/019 Control Rod Drive System 036

OP/1/A/1502/007 Ops Defueling Responsibilities 093

OP/1/A/1502/009 Containment Closure Control 052

PT/0/A/0711/001 Zero Power Physics Test 077

PT/0/A/0750/012 Development of Fuel Movement Instructions Procedure 052

PT/0/A/0750/018 Refueling Activities 026

PT/1/A/0115/012 Unborated Water Source Isolation 005

PT/1/A/0630/001 Mode Change Verification 039

Work Orders 20513301, 20513319, 20513852

71111.22 Corrective Action 2398057, 2374879, 2378213, 2402552, 2413125, 2413595,

Documents 2328195, 2447233, 2447755, 2447066

Drawings OFD-101A-1.1 Flow Diagram of High Pressure Injection System (Letdown 53

Section)

OFD-101A-1.2 Flow Diagram of High Pressure Injection System (Storage 47

Section)

OFD-101A-1.3 Flow Diagram of High Pressure Injection System (Charging 37

Section)

OFD-101A-1.4 Flow Diagram of High Pressure Injection System (Charging 51

Section)

OFD-104A-1.1 Flow Diagram of Spent Fuel Cooling System 68

ONTC-1-LP-002 Test Acceptance Criteria for Valves 1LP-2 19

Inspection Type Designation Description or Title Revision or

Procedure Date

Miscellaneous Inservice Test (IST) Acceptance Criteria Procedure

Datasheets, 1HPI-PU-0001/2/3, on November 18, 2022

Safe Industries Quarterly PM Inspection/Maintenance

Checklist on October 26, 2022

AD-EG-ALL-1450 Preconditioning of Structures, Systems, and Components 1

CSD-EG-ONS-Diverse and Flexible Coping Strategies (FLEX) Program 005

1619.1000 Document - Oconee Nuclear Station

OSS-0254.00-00- (Mech) High Pressure Injection and Purification & 062

1001 Deborating Demineralizer Systems

Procedures IP/0/A/0050/008 FLEX Building Equipment Inventory 001

OP/1/A/1104/002 HPI System 182

PT/1/A/0150/034 Leak Rate History Record and Reactor Building Leak Rate 014

Verification

PT/1/A/0151/056 Penetration 56 Leak Rate Test 007

PT/1/A/0251/024 HPI Full Flow Test 054

Work Orders 20199103, 20218705, 20547704, 20513777, 20513141,

20528609, 20432973

71114.06 Miscellaneous EP-ALL-EPLAN Duke Energy Common Emergency Plan 3

71152A Corrective Action PIRT 2443366

Documents 2429285, 2443672

Drawings ONS-M-Survey of Room 254 Unit 3 SFP Demineralizer and Filters 07/09/2022

220709-1

Miscellaneous WR 20207276, 20227289

Oconee Dose Rate Alarm Report, September 2022

Procedures AD-PI-ALL-0100 Corrective Action Program 27

AD-PI-ALL-0100 Corrective Action Program 28

AD-RP-ALL-0004 Radiological Posting and Labeling 6

AD-RP-ALL-0005 Access Controls for High and Locked High Radiation Areas 1

AD-RP-ALL-2017 Access Controls to Very High Radiation Areas and 8

Supplemental Access Controls for HRA and LHRA

HP/0/B/1000/054 Radiation Protection Routines 053

Work Orders 20561193

71152S Corrective Action 2185215

Documents

Inspection Type Designation Description or Title Revision or

Procedure Date

Miscellaneous WR 20099106

2