IR 05000269/2022004
| 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 Aspect Report Section Initiating Events Green NCV 05000269/2022004-01 Open/Closed
[H.12] - Avoid Complacency 71111.04 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 Aspect Report Section Mitigating Systems Green NCV 05000269,05000270,05000287/2022004-02 Open/Closed None (NPP)71111.15 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 Aspect Report Section Occupational Radiation Safety Green NCV 05000287/2022004-03 Open/Closed
[H.12] - Avoid Complacency 71152A 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:
1. 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)
2. 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
3. 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
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 Aspect Report Section Initiating Events Green NCV 05000269/2022004-01 Open/Closed
[H.12] - Avoid Complacency 71111.04 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 Aspect Report Section Mitigating Systems Green NCV 05000269,05000270,05000287/20220 04-02 Open/Closed None (NPP)71111.15 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. Specifically, 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 Aspect Report Section Occupational Radiation Safety Green NCV 05000287/2022004-03 Open/Closed
[H.12] - Avoid Complacency 71152A 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 inspectors 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 without 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
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
2164850, O-08-01643, 2396702, 2366042, 2373308,
2449312, 2451863, 2450272, 2438431, 2450253, 2451877,
26303
Drawings
O-0702
One Line Diagram 6900V & 4160V STA Auxiliary Sys
O-0702-A
One Line Diagram 6900V & 4160V Auxiliary Sys
O-0703-G
One Line Diagram Station Auxiliary Circuits 600/208V L/C
1X8, 1X9, 1X10 & MCC 1XS1, 1XS2, 1XS3
O-703-E
One Line Diagram Station Auxiliary Circuits 600V/208V L/C
1X6 & MCC 1XI, 1XN, 1XP & 1XQ
OFD-101A-1.1
Flow Diagram of High Pressure Injection System Letdown
Section
OFD-101A-1.2
Flow Diagram of High Pressure Injection System Storage
Section
OFD-101A-1.3
Flow Diagram of High Pressure Injection System Charging
Section
OFD-101A-1.4
Flow Diagram of High Pressure Injection System Charging
Section
OFD-101A-1.5
Flow Diagram of High Pressure Injection System SSF
Portion
OFD-130A-2.1
Flow Diagram of Essential Siphon Vacuum System
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
2
Work Orders
20513544, 20518531, 20285643, 20406681, 20237935
Calculations
OSC-9314
NFPA 805 Transition Risk-Informed Performance-Based
Fire Risk Evaluation
006
Corrective Action
Documents
274272, 2407000
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Fire Plans
CSD-ONS-FS-
019
Standard Operating Guide Fire Inside Containment
000
CSD-ONS-PFP-
Pre-Fire Plan for U1 Auxiliary Building Elevation 796
001
CSD-ONS-PFP-
1RB
Pre-Fire Plan for U1 Reactor Building
000
CSD-ONS-PFP-
Pre-Fire Plan for U2 Auxiliary Building Elevation 809
CSD-ONS-PFP-
Pre-Fire Plan for U2 Turbine Building Elevation 796
000
CSD-ONS-PFP-
Pre-Fire Plan for U3 Auxiliary Building Elevation 783
O-0310-FZ-009
Auxiliary Building Unit 1 Fire Protection Plan Fire Area & Fire
Zone Boundaries Plan at EL 796+6 & EL 797+6
O-0310-FZ-029
Turbine Building Unit 2 Fire Protection Plan Fire Area & Fire
Zone Boundaries Plan at Mezzanine EL 796+6
O-0310-K-007
Fire Protect Aux Bldg Unit 1 EL 796+6
O-0310-L-005
Turbine Building Unit 2 Fire Protection Plan & Fire Barrier,
Flood, & Pressure Boundaries Plan at Mezzanine EL 796+6
O-310-K-06
Auxiliary Building - Unit 3 Fire Protection Plan & Fire Barrier,
Flood, & Pressure Boundaries Plan at EL 783+9
O-310-K-11
Auxiliary Building - Unit 2 Fire Protection Plan & Fire Barrier,
Flood, & Pressure Boundaries Plan at EL 809+3
Miscellaneous
AD-FP-ALL-1520
Transient Combustible Control
AD-WC-ONS-
20
ONS Shutdown Risk Management
OFD-124C-1.2
Flow Diagram of High Pressure Service Water System
Turbine Building
Work Orders
239781, 20239784
Calculations
OSC-2693
Decay Heat Removal Cooler Performance Calculation
Methodology Verification
OSC-3993
U1 LI Heat Exchanger Performance Calculation
Drawings
OM 201-0286.001
- 46-128:3940 Decay Heat Coolers
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
IP/1/A/0203/001 C
Unit 1 Low Pressure Injection System Accessible Flow
Instruments Calibration
PT/1/A/0251/069
LPI Cooler Test
Corrective Action
Documents
NCRs (by number)
2436679, 02447643, 02447680, 02447967
NDE Reports
PT-22-001
Liquid Penetrant Examination of RPV Head Forging
Surface
10/31/2022
UT-22-017
Ultrasonic Examination of 1RC-201-121 (PDI-UT-3 for flaw
disposition)
11/01/2022
UT-22-024
Ultrasonic Examination of 1RC-201-121
10/31/2022
Procedures
OP-OC-15JT-05
Night Shift Shutdown JITT
2a
OP/1/A/1102/010
Controlling Procedure for Unit Shutdown
236
Calculations
OSC-7966
EWST Inventory for Turbine Building Flood
010
OSC-8729
Supporting Vendor Analysis for Alloy 600 Mitigation -
Oconee Nuclear Station Units 1, 2 & 3 - HPI Evaluations
Supporting Environmental Fatigue Effects
D2
Corrective Action
Documents
2445019, 2326989, 2336744, 2342922, 2431932
Drawings
OFD-124C-1.4
Flow Diagram of High Pressure Service Water System
West Yard
OFD-124C-1.5
Flow Diagram of High Pressure Service Water System East
Yard and Visitors Center
OM 201-597.001
Assembly & Detail for 2-1/2" Pressure Injection Nozzle
D2
Engineering
Changes
111894
QA-1 Weld Build Up on 1B2 Nozzle Safe End
21842
QA-1 Weld Overlay on 1B2 HPI Nozzle Safe End to Pipe
Miscellaneous
AD-EG-ALL-1210
Structural Integrity
Associates, Inc.
Report
201209.310
Oconee Unit 1 HPI Design Reconciliation & Qualification
Structural Integrity
Associates, Inc.
Report
201209.401
Interim Assessment of Crack Growth - HPI Nozzle 1B2
Weld Overlay ONS1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Structural Integrity
Associates, Inc.
Report
201209.402
ONS1 1B2 HPI Nozzle - Safe End to Pipe Weld Overlay
Repair Compliance Matrix for ASME Code,Section XI,
Nonmandatory Appendix Q, 2017 Edition
TE-MN-ALL-0510
Excavation and Backfill
Procedures
MP/0/A/1705/008 A
Fire Protection - Hydrant - Disassembly, Repair, and
Reassembly
2
OP/0/A/1104/011
High Pressure Service Water
108
OP/1/A/6101/009
Alarm Response Guide 1SA-09
2
Work Orders
20513217, 20418874, 20517775
Calculations
OSC-1306.06
Piping Analysis for Low Pressure Service Water System to
and from RBCU Cooling Coils
D13
OSC-4553
Maximum Differential Pressures for LPSW Valves
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
Valves at Oconee
Corrective Action
Documents
2449176, 2386632
Drawings
OFD-124B-1.2
Flow Diagram of Low Pressure Service Water System
Reactor Building Cooling Units
OM 202-0004.001
80X96 Type SAFV Pump
ONTC-1-LPSW-139 Test Acceptance Criteria for Valve 1LPSW-139
Miscellaneous
ONEI-0400-400
Revised RBCU Requirements in Support of 1-Year Mission
Times
Work Orders
20568729, 20398370, 20566826, 20566827
Corrective Action
Documents
2430827, 2426484
Drawings
O-422M-32
Instrument Detail Startup FDW Valve Control 1FDW-35
& 1FDW-44
OFD-121B-1.3
Flow Diagram of Feedwater System (Final Feedwater)
Engineering
Changes
21202
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
IP/1/A/0275/012 A
Unit 1 Feedwater Control Valve Demand and Interlock
Calibration
004
Work Orders
20542872
Corrective Action
Documents
1761227, 1794841, 2448302, 2450428, 2450570
Drawings
O-422BB-6
Instrument Detail Reactor Vessel Level Loop A (1RC
LT0005A)
O-422BB-6.01
Instrument Detail Reactor Vessel Level Loop B (1RC
LT0005B)
OFD-104A-1.1
Flow Diagram of Spent Fuel Cooling System
OFD-104A-1.2
Flow Diagram of Spent Fuel Cooling System Purification
Loop
Engineering
Changes
21670
Allowable Alternate Power Configuration for a SFP Cooling
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 Specification 3.7.13 - Definitions, Boundary Conditions, and
Spent Fuel Pool Storage Qualifications
ONEI-0400-575
Oconee 1 Cycle 33 Final Core Load Map
ONEI-0400-583
Oconee 1 Cycle 33 Core Operating Limits Report
ONEI-0400-584
O1C33 Physics Test Manual
Procedures
AD-MN-ALL-0002
AD-SY-ALL-0460
Managing Fatigue and Work Hour Limits
AD-WC-ALL-0420
Shutdown Risk Management
AD-WC-ONS-
20
ONS Shutdown Risk Management
AP/1-
2/A/1700/035
Loss of SFP Cooling and/or Level
AP/1/A/1700/026
Loss of Decay Heat Removal
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
IP/0/B/0200/027 A
Reactor Level Instrumentation and Calibration
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
OP/1/A/1102/001
Controlling Procedure for Unit Startup
25
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
2
PT/0/A/0711/001
Zero Power Physics Test
077
PT/0/A/0750/012
Development of Fuel Movement Instructions Procedure
2
PT/0/A/0750/018
Refueling Activities
26
PT/1/A/0115/012
Unborated Water Source Isolation
005
PT/1/A/0630/001
Mode Change Verification
039
Work Orders
20513301, 20513319, 20513852
Corrective Action
Documents
2398057, 2374879, 2378213, 2402552, 2413125, 2413595,
28195, 2447233, 2447755, 2447066
Drawings
OFD-101A-1.1
Flow Diagram of High Pressure Injection System (Letdown
Section)
OFD-101A-1.2
Flow Diagram of High Pressure Injection System (Storage
Section)
OFD-101A-1.3
Flow Diagram of High Pressure Injection System (Charging
Section)
OFD-101A-1.4
Flow Diagram of High Pressure Injection System (Charging
Section)
OFD-104A-1.1
Flow Diagram of Spent Fuel Cooling System
ONTC-1-LP-002
Test Acceptance Criteria for Valves 1LP-2
Inspection
Procedure
Type
Designation
Description or Title
Revision or
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
CSD-EG-ONS-
1619.1000
Diverse and Flexible Coping Strategies (FLEX) Program
Document - Oconee Nuclear Station
005
OSS-0254.00-00-
1001
(Mech) High Pressure Injection and Purification &
Deborating Demineralizer Systems
2
Procedures
IP/0/A/0050/008
FLEX Building Equipment Inventory
001
OP/1/A/1104/002
HPI System
2
PT/1/A/0150/034
Leak Rate History Record and Reactor Building Leak Rate
Verification
014
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
Miscellaneous
Duke Energy Common Emergency Plan
Corrective Action
Documents
PIRT 2443366
29285, 2443672
Drawings
ONS-M-
220709-1
Survey of Room 254 Unit 3 SFP Demineralizer and Filters
07/09/2022
Miscellaneous
WR 20207276, 20227289
Oconee Dose Rate Alarm Report, September 2022
Procedures
AD-PI-ALL-0100
Corrective Action Program
AD-PI-ALL-0100
Corrective Action Program
AD-RP-ALL-0004
Radiological Posting and Labeling
AD-RP-ALL-0005
Access Controls for High and Locked High Radiation Areas
AD-RP-ALL-2017
Access Controls to Very High Radiation Areas and
Supplemental Access Controls for HRA and LHRA
HP/0/B/1000/054
Radiation Protection Routines
053
Work Orders
20561193
Corrective Action
Documents
2185215
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous