IR 05000335/2022003

– Integrated Inspection Report 05000335/2022003 and 05000389/2022003

ML22308A067
Person / Time
Site:	Saint Lucie
Issue date:	11/04/2022
From:	David Dumbacher NRC/RGN-II/DRP/RPB3
To:	Coffey B Florida Power & Light Co
References
IR 2022003
Download: ML22308A067 (24)

Text

November 04, 2022

SUBJECT:

ST. LUCIE UNITS 1 & 2 - INTEGRATED INSPECTION REPORT 05000335/2022003 AND 05000389/2022003

Dear Bob Coffey:

On September 30, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at St. Lucie Units 1 & 2. On October 27, 2022, the NRC inspectors discussed the results of this inspection with Mr. Carlos Santos, Operations Director, and other members of your staff. The results of this inspection are documented in the enclosed report.

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violation or the significance or severity of the violation 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 St. Lucie Units 1 & 2.

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 St. Lucie Units 1 & 2. 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, David E. Dumbacher, Chief Reactor Projects Branch 3 Division of Reactor Projects Docket Nos. 05000335 and 05000389 License Nos. DPR-67 and NPF-16

Enclosure:

NRC Inspection Report (IR) 05000335/2022003 05000389/2022003 w/attachment: IMC 0609, Appendix M

Inspection Report

Docket Numbers: 05000335 and 05000389

License Numbers: DPR-67 and NPF-16

Report Numbers: 05000335/2022003 and 05000389/2022003

Licensee: Florida Power & Light Company

Facility: St. Lucie Units 1 & 2

Location: Jensen Beach, FL

Inspection Dates: July 01, 2022 to September 30, 2022

Inspectors: M. Magyar, Reactor Inspector D. Orr, Senior Resident Inspector B. Pursley, Health Physicist R. Reyes, Resident Inspector J. Rivera, Health Physicist S. Roberts, Resident Inspector S. Sandal, Senior Reactor Analyst M. Schwieg, Senior Reactor Inspector

Approved By: David E. Dumbacher, Chief Reactor Projects Branch 3 Division of Reactor Projects

Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at St. Lucie Units 1 & 2, 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

Reactor Shutdown Required by Technical Specifications due to a Misaligned Control Element Assembly (CEA)

Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.8] - 71111.12 Systems NCV 05000389/2022003-01 Procedure Open/Closed Adherence A self-revealed Green NCV of TS 6.8.1.b., Procedures and Programs, was identified for failure to implement refueling operations procedure 0-NOP-67.11, CEA Extension Shaft Coupling and Uncoupling Using Hydraulic Operated Tool (SCOUT), and failure to follow FME procedure requirements to perform a post-service inspection of a SCOUT tool during the St.

Lucie unit 2 fall 2021 refueling outage.

Additional Tracking Items

None.

PLANT STATUS

Unit 1 operated at or near rated thermal power (RTP) until the unit was shut down for planned refueling outage, SL1-31, on September 2, 2022. Unit 1 remained in SL1-31 when the inspection period ended on September 30, 2022.

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

INSPECTION SCOPES

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

71111.01 - Adverse Weather Protection

Impending Severe Weather Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the adequacy of the overall preparations to protect risk significant systems from impending severe weather, Hurricane Ian, on September 27 - 29, 2022.

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 intake cooling water (ICW) system while the 1B ICW pump was out of service (OOS) for an emergent repair on July 18, 2022

(2) 1B fire water pump while the 1A fire water pump was OOS for maintenance on August 17, 2022

(3) Unit 1 shutdown cooling system while in operation for decay heat removal on September 7, 2022

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

(1) The inspectors evaluated system configurations during a complete walkdown of the unit 1 high pressure safety injection (HPSI) system on September 8, 2022.

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 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) Unit 1 fire zone (FZ) 5, component cooling water (CCW) area on July, 21, 2022

(2) Unit 2 FZs 22 and 23, A and B electrical penetration rooms on July 27, 2022

(3) Unit 2 FZ 3, CCW building on July 27, 2022

(4) Unit 1 FZ 45, piping penetration room on August 20, 2022

(5) Unit 1 FZ 33, pipe tunnel on August 23, 2022

(6) Unit 1 FZ 26, containment building on September 22, 2022

71111.06 - Flood Protection Measures

Inspection Activities - Internal Flooding (IP Section 03.01) (2 Samples)

The inspectors evaluated internal flooding mitigation protections in the:

(1) Unit 1 manhole 156 on July 19, 2022

(2) Unit 2 reactor auxiliary building (RAB) elevation -0.5 feet on August 10, 2022

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) Alternate spent fuel pool (SFP) cooling system aligned to the unit 1 SFP on September 12, 2022

71111.07T - Heat Exchanger/Sink Performance Heat Exchanger (Service Water Cooled) (IP Section 03.02)

The inspectors evaluated heat exchanger performance on the following:

(1) 1A CCW heat exchanger

Heat Exchanger (Closed Loop) (IP Section 03.03) (1 Sample)

The inspectors evaluated heat exchanger performance on the following:

(1) 2A and 2B shutdown cooling heat exchangers

Ultimate Heat Sink (IP Section 03.04) (1 Sample)

The inspectors evaluated the ultimate heat sink performance on the following:

(1) Big Mud Creek, Ultimate Heat Sink

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

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

(1) The inspectors evaluated pressurized water reactor non-destructive testing by reviewing the following examinations from September 12 - September 15, 2022:

1. Ultrasonic Testing a. CS-1-1-SW-2, Pipe to Elbow, Class 1 (observed)b. CS-1-1-SW-6, Pipe to Elbow, Class 1 (observed)c.

SI-112-FW-2000, Reducer to Pipe, Class 1 (observed)d. SI-112-FW-2002, Tee to Valve V3124, Class 1 (observed)e. SI-142-FW-2000, Pipe to Valve V3124, Class 1 (reviewed)f.

SI-142-FW-2001, Pipe to Pipe, Class 1 (observed)

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

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:

A planned power reduction and shut down of unit 1 to support refueling outage SL1-31 on September 2-3, 2022 Unit 1 solid plant pressure control on September 23, 2022

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

(1) The inspectors observed and evaluated a licensed operator continuing training evaluation in the control room simulator on July 25, 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) AR 2432144, 1B ICW pump tripped on overcurrent, review completed on September 9, 2022

(2) AR 2415359, during 2-OSP-66.01, CEA 27 slipped in causing greater than 15-inch deviation, review completed on July 31, 2022

Quality Control (IP Section 03.02) (1 Sample)

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

(1) Work order (WO) 40692292, V3124, check valve for low pressure safety injection (LPSI) feed to reactor coolant system loop 1A1, replacement on September 26, 2022

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)

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

(1) Unit 2 elevated risk when the 2B LPSI train was OOS for planned maintenance on July 28, 2022

(2) Units 1 and 2 elevated risk when the 1B and 2B startup transformers were OOS for planned maintenance on the 2B4 switchgear of August 22, 2022

(3) Unit 2 elevated risk when the 2B startup transformer was OOS for planned maintenance on September 12, 2022

71111.15 - Operability Determinations and Functionality Assessments

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

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

(1) AR 2432049, B AFAS SG-1 Delta-P Pretrip Light Did Not Illuminate, review completed on July 22, 2022

(2) AR 2432163, Unit 2 Annunciator G-25 AFAS Cabinet Trouble Locked In, review completed on July 26, 2022

(3) AR 2429763, 1B2 Engine Coolant Leak, review completed on August 3, 2022

(4) AR 2434532, Unit 1 Reactor Protective System Channel C, Thermal Margin/Low Pressure Trip Bistable Did Not Reset, review completed on August 18, 2022

(5) AR 2434692, PT-0801A & 1B, SG 2A & 2B Main Steam Header Pressure Transmitters Past EQ End of Life, review completed on August 24, 2022

71111.18 - Plant Modifications

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

The inspectors evaluated the following temporary or permanent modifications:

(1) EC 297558, freeze seal to support replacing V3124, check valve for LPSI feed to reactor coolant system loop 1A1, on September 12, 2022

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (8 Samples)

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

(1) WO 40270260, 1B ICW pump motor replacement, reviewed on July 26, 2022

(2) WO 40772147, U2 HCV-3625, LPSI flow control valve operator, open limit switch setting verification, reviewed on August 2, 2022

(3) WO 40680353, 2A auxiliary feedwater (AFW) pump breaker replacement, reviewed on August 10, 2022

(4) WO 40789606, 1B3 4KV switchgear breaker replacements, reviewed on September 26, 2022

(5) WO 40692292, V3124, check valve for LPSI feed to reactor coolant system loop 1A1, replacement, reviewed on September 27, 2022

(6) WO 40469218, 1B EDG voltage regulator and excitation system replacement, reviewed on September 22, 2022

(7) WO 40627719, 125Vdc 1B battery replacement, reviewed on September 22, 2022

(8) WO 40641402, MV-09-9, 1A AFW pump flow control valve to 1A steam generator, stem/plug replacement, reviewed on September 30, 2022

71111.20 - Refueling and Other Outage Activities

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

(1) (Partial)

The inspectors evaluated refueling outage SL1-31 activities from September 3, 2022, to the end of the inspection period.

71111.22 - Surveillance Testing

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

Surveillance Tests (other) (IP Section 03.01) (3 Samples)

(1) 0-OSP-37.01, Emergency Cooling Water Canal - Periodic Test, reviewed on July 26, 2022

(2) 2-OSP-59.01B, 2B Emergency Diesel Generator Monthly Surveillance, reviewed on August 20, 2022

(3) 1-SMM-08.08, Main Steam Safety Valve Setpoint Surveillance Using Furmanite Trevitest Mark VIII Equipment, reviewed on September 9, 2022

Inservice Testing (IP Section 03.01) (2 Samples)

(1) 2-OSP-14.02A, 2A Component Cooling Water Pump Comprehensive Pump Test, reviewed on August 11, 2022

(2) 2-OSP-03.05B, 2B High Pressure Safety Injection Pump Code Run, reviewed on August 19, 2022

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

(1) 1-OSP-68.02, Local Leak Rate Test, Penetration 54 - Integrated Leak Rate Test (ILRT) Pressurization Station, reviewed on September 22, 2022

FLEX Testing (IP Section 03.02) (1 Sample)

(1) 0-OSP-83.04, Periodic Testing of the Flex 480V Diesel Generators, Units B and C, reviewed on July 7,

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards during the unit 1 refueling outage number 31 (SL1-31).

Instructions to Workers (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards during SL1-31.

Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)

The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material during SL1-31:

(1) Licensee surveys of potentially contaminated material leaving the radiological control area (RCA) at the south service building RCA control point

(2) Licensee surveys and control of potentially contaminated material going in and out of the unit 1 reactor containment building

(3) Workers exiting the reactor containment building and the south service building exit RCA

Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)

The inspectors evaluated the licensee's control of radiological hazards for the following radiological work during SL1-31:

(1) Radiation Protection Plan for Valve 3124 Replacement Project

(2) Radiation Protection Plan SL1-31 for Replacement CEDM Coil Stacks and RSPTs

(3) RWP 22-1005, Temp Rx Head: Install/Remove/Decon Temp Rx Head

(4) RWP 22-1410, Code Safeties, PORV's, PZR Manway, PZR Miscellaneous Valve Work High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (4 Samples)

The inspectors evaluated licensee controls of the following High Radiation Areas and Very High Radiation Areas during SL1-31:

(1) Locked High Radiation Area (LHRA) at access to unit 1 reactor containment regen heat exchanger room.

(2) LHRA access at unit 1 reactor containment reactor head keyway.

(3) LHRA access at unit 1 reactor containment building blast damper doors.

(4) LHRA access at unit 1 reactor containment building 62' lower Cavity Ladder Lock.

Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)

(1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements during SL1-31.

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

Transportation

Radioactive Material Storage (IP Section 03.01)

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

(1) Dry Storage Warehouse

(2) RCA Yard

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

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

(1) Radwaste Processing Skid in U1 RAB Drumming Station Room

Waste Characterization and Classification (IP Section 03.03) (3 Samples)

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

(1) 2B CVCS Filter Waste Stream

(2) U1 SFP Purification Filter Waste Stream

(3) Dry Active Waste Stream

Shipment Preparation (IP Section 03.04) (1 Sample)

(1) no radioactive shipments available to observe during the inspection

Shipping Records (IP Section 03.05) (4 Samples)

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

(1) Shipment no. 2021-43, Resin Liner, LSA II

(2) Shipment no. 2021-100, Filter Liner, Type B

(3) Shipment no. 2020-57, Dry Active Waste, LSA II

(4) Shipment no. 2022-9, Tools and Equipment, SCO II

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===

(1) Unit 1, July 1, 2021 through June 30, 2022

(2) Unit 2, July 1, 2021 through June 30, 2022

OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)

(1) October 31, 2021 through August 31, 2022

PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)

(1) August 18, 2021 through September 6, 2022

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 issues:

(1) External corrosion under insulation inspection program implementation documented in ARs 2431068, 2432399, 2433128, 2434637, and completed in WO 40844554 during the SL1-31 refuel outage.

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 equipment issues that resulted in operational mode holds during the SL1-31 refuel outage that might be indicative of a more significant safety issue.

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Followup (IP Section 03.01)

(1) On September 2, 2022, at 2248 hours0.026 days <br />0.624 hours <br />0.00372 weeks <br />8.55364e-4 months <br />, St. Lucie Unit 1 reactor was manually tripped from 40% RTP in response to lowering steam generator water levels. An auxiliary feedwater actuation signal occurred due to lowering steam generator levels. Control room operators were in the process of down powering unit 1 in preparation for refueling outage SL1-31. Operators, by procedure, planned to manually trip unit 1 when 25% RTP was established, but at about 40% RTP, the main condenser hotwell recirculation spray header flow control valve failed open, reducing main feedwater pump suction pressure and the single operating main feedwater pump tripped on low suction pressure. The resident inspector was on-site observing the unit 1 shutdown and verified that stable plant conditions were established, and plant systems responded as designed.

Event Report (IP Section 03.02) (1 Sample)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 05000389/2022-001-00, Reactor Shutdown Required by Technical Specifications Due to a Misaligned Control Element Assembly (ADAMS Accession No. ML22063A075). The inspection conclusions associated with this LER are documented in this inspection report under Inspection Results Section 71111.12, Maintenance Effectiveness.

INSPECTION RESULTS

Reactor Shutdown Required by Technical Specifications due to a Misaligned Control Element Assembly (CEA)

Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.8] - 71111.12 Systems NCV 05000389/2022003-01 Procedure Open/Closed Adherence A self-revealed Green NCV of TS 6.8.1.b., Procedures and Programs, was identified for failure to implement refueling operations procedure 0-NOP-67.11, CEA Extension Shaft Coupling and Uncoupling Using Hydraulic Operated Tool (SCOUT), and failure to follow FME procedure requirements to perform a post-service inspection of a SCOUT tool during the St.

Lucie unit 2 fall 2021 refueling outage.

Description:

On January 6, 2022, at 0824 hours0.00954 days <br />0.229 hours <br />0.00136 weeks <br />3.13532e-4 months <br />, surveillance testing in accordance with procedure 2-OSP-66.01, CEA Exercise, was being performed on unit 2 CEAs. During CEA 27 exercising, CEA 27 slipped from 133 inches to 120 inches. The condition resulted in a position deviation greater than 15 inches from all other CEAs in its group and TS action statement 3.1.3.e. applied. Subsequent attempts to recover CEA 27, while at 70% power and within the one-hour time limitation of TS action statement 3.1.3.e., were not successful and unit 2 was shut down to hot standby to comply with TS requirements. The event was described in LER 05000389/2022001 Revision 0, dated March 4, 2022 (ADAMS Accession Number ML22063A075).

The licensee performed troubleshooting using its failure investigation process and determined a malfunction of the associated control element drive mechanism (CEDM) was the likely cause of CEA 27 slipping. The licensee cooled down unit 2 to cold shutdown and replaced CEDM 27.

After removal of the malfunctioned CEDM 27 from its housing on the reactor head, an inspection was performed using Westinghouse inspection guidelines. The inspection identified a 0.25-inch diameter and 0.3125-inch-long cylindrical metallic object attached to the CEDM motor. The object was between the upper grippers moveable latch magnet and the top of the pull-down magnet. The location of the object between these components interfered with full travel of the upper gripper moveable latch during CEDM operation. It was suspected that the interference prevented the grippers from closing on the notched areas of the CEA extension shaft resulting in CEA 27 slipping. The suspected behavior was also consistent with the abnormal CEDM current traces that were observed and recorded on January 6, 2022.

CEDM 27 motor was packaged and shipped to a Westinghouse facility for further inspection, disassembly, and analysis to determine the source of the FME found lodged between the upper grippers moveable latch magnet and the top of the pull-down magnet. Subsequently, the SCOUT tool that was used to latch CEA 27 during the most recent fall 2021 refuel outage, was shipped to a Westinghouse facility for disassembly and inspection. There it was determined that two pins were missing from the latching mechanism of the SCOUT tool. The object found in CEDM 27 was examined and determined to be consistent in size, geometry, and material to one of two L-slot pins in the SCOUT tool. The second L-slot pin was not retrieved from the site.

As it could not be confirmed that the missing L-slot pin is not present in the reactor coolant system (RCS), Westinghouse performed a loose parts evaluation (LPEV), CSTL2-LPEV-TR-MZ-000001, Revision A-0, St Lucie Unit 2 Loose Parts Evaluation, to examine the potential effects of subsequent Unit 2 operation on RCS components, major primary system components, as well as auxiliary systems assuming the missing L-slot pin remained within the Unit 2 primary system at the conclusion of the Unit 2 fall 2021 refuel outage. The Westinghouse LPEV also considered a missing SCOUT tool handle roll pin but subsequent to the LPEV, Westinghouse and the licensee were able to determine the SCOUT tool handle roll pin was missing prior to the Unit 2 refuel outage. The licensee also consulted with the reactor fuel assembly vendor to assess the potential impact of the missing L-slot pin on the fuel assemblies should the L-slot pin remain intact or fragment.

The Westinghouse LPEV concluded that the missing L-slot pin was bounded by parts that have been previously evaluated at other Combustion Engineering and Westinghouse-designed plants. The LPEV concluded that the operation of St Lucie Unit 2 for one cycle of operation following the Spring 2022 forced outage with the missing pins is acceptable. The inspectors noted in the Westinghouse LPEV, that it is possible for one CEA to be prevented from rapidly inserting. The LPEV noted that analyses have shown that the reactor can be safely shutdown with the highest worth CEA stuck in the fully withdrawn position.

Framatome LPEV, FS1-0062463, Revision 1, SLU2-26 (and Future Cycles) Loose Parts Assessment, evaluated the risk and potential impact on fuel assembly performance based on the L-slot pin characterization. The Framatome LPEV concluded there is a small possibility for the L-slot pin to cause a debris related fuel failure but monitoring of the core chemistry would indicate if a debris failure has occurred and allow further evaluation of the fuel assemblies and core. The Framatome LPEV additionally concluded that the CEAs on the current cycle were inserted prior to the L-slot pin being lost, therefore, safe shutdown of the reactor is not a concern. The missing L-slot pin also would also not prevent CEA insertion or shutdown, would not sufficiently block coolant flow to affect the thermal hydraulic performance of a fuel assembly, and the L-slot pin in its entirety or fragmented, has minimal potential to promote corrosion.

The licensee completed a root cause evaluation in AR 2415359 and determined that FME program requirements for complex tools were not adequately followed. The reactor services crew's failure to follow FME program requirements for the SCOUT tool prevented the discovery of the L-slot pins being missing from the SCOUT tool during CEA coupling. Specifically, procedure 0-NOP-67.11, CEA Extension Shaft Coupling and Uncoupling Using Hydraulic Operated Tool (SCOUT), step 4.1, Initial Activities, requires that Section 2.0, Precautions and Limitations has been reviewed. Precaution 2.1.1 states the FME area requirements of MA-AA-101-1000, Foreign Material Exclusion Procedure, shall be followed during the performance of 0-NOP-67.11. MA-AA-101-1000, requires that MA-AA-101-1000-F08, FME Complex Equipment and Tooling Inspection Checklist, to be completed when complex tooling, such as the SCOUT tool, are used in the performance of work.

Had MA-AA-101-1000-F08 been used during CEA coupling on September 16, 2021, detailed pre, periodic, and post-use inspections of the SCOUT tool would have been performed and documented. Although the MA-AA-101-1000-F08 checklist was not used during CEA coupling, a pre-inspection of the SCOUT tool was performed. The pre-inspection was documented in WO 40717564, U2 CEA Latching Tool; Inspect and Repair. The WO did not document any deficiency with the tool prior to its use. During CEA 27 coupling, the reactor services crew had difficulty disengaging the SCOUT tool from the CEA extension shaft L-slot. The tool was shaken free, and it is assumed that one of the two L-slot pins failed at that time. The CEA coupling activities continued with the same SCOUT tool after completing CEA 27. The SCOUT tool will function with only one L-slot pin. The crew was unaware that damage had occurred to the SCOUT tool. The pin would remain on the CEA 27 extension shaft, through magnetism, until reactor head installation, when it fell off as the shaft passed through the CEDM. The coupling activities were completed for an additional forty CEAs without issue. When attempting to couple CEA 24, the reactor services crew could not engage the tool L-slot pins. It is assumed the remaining L-slot pin failed at that point. The tool was replaced, a post-inspection of the failed SCOUT tool was not performed, and the remaining CEA extension shafts were coupled with the new SCOUT tool. The reactor services crew missed an opportunity to identify the missing L-slot pins by not performing a post-use inspection or questioning the difficulty in removing the SCOUT tool from the CEA 27 extension shaft.

Corrective Actions: The licensee addressed the causal factors with the performance issues that resulted in 1) FME that was retained within CEDM 27; and 2) a second L-slot pin potentially lost within the unit 2 RCS. The root cause evaluation, causal factors, and corrective actions were documented in AR 2415359. Corrective actions included replacing CEDM 27, incorporating complex tool inspection into the reactor services FME plan, revising procedure 0-NOP-67.11 to address FME risks during CEA coupling, and work order activities to perform a foreign object search and retrieval activity for the missing L-slot pin during the next unit 2 refueling outage.

Corrective Action References: AR 2415359

Performance Assessment:

Performance Deficiency: The licensees failure to perform a post-use inspection of the SCOUT tool after it failed and was replaced by another SCOUT tool during CEA extension shaft coupling, as required by MA-AA-101-1000-F08, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a potential loose part has been evaluated to exist within another CEDM and may prevent its ability to rapidly insert.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that more than one cornerstone was affected by the finding and associated foreign material condition in the unit 2 RCS. Therefore, the significance determination of this finding considered the impact to the mitigating systems and barrier integrity cornerstones.

For the mitigating systems cornerstone, the inspectors noted that it was possible for one CEA to be prevented from rapidly inserting and that analyses have shown that the reactor can be safely shut down with the highest worth CEA stuck in the fully withdrawn position. Therefore, using Exhibit 2 - Mitigating Systems Screening Questions of IMC 0609 Appendix A, this issue screened as Green because the finding only affected a single reactor protection system (RPS) trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown (e.g., other automatic RPS trips, alternate rod insertion, or manual reactor trip capacity) were maintained.

For the barrier integrity cornerstone, the inspectors used Exhibit 3 - Barrier Integrity Screening Questions, and determined this issue required further evaluation using Appendix M, Significance Determination Process Using Qualitative Criteria, because the finding was the result of mismanagement of the FME or reactor coolant chemistry control program that challenged fuel cladding integrity (e.g., loose parts, material controls). An Appendix M evaluation was completed by a Senior Reactor Analyst (SRA) using Table 1, Qualitative Decision-Making Attributes for NRC Management Review of IMC 0609 Appendix M. The evaluation considered each attribute in Table 1 and concluded this performance deficiency was Green because of the low mass of material involved, the small potential for isolated fuel cladding damage, and that any potential future fuel cladding damage would be readily detectable through primary chemistry monitoring programs.

Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. The MA-AA-101-1000-F08 checklist was not used during CEA coupling and detailed pre, periodic, and post-use inspections of the SCOUT tool were not performed and documented.

Enforcement:

Violation: St. Lucie Unit 2 TS 6.8.1.b., states, in part, that written procedures shall be established, implemented, and maintained covering refueling operations. Procedure 0-NOP-67.11, CEA Extension Shaft Coupling and Uncoupling Using Hydraulic Operated Tool (SCOUT), is a refueling operations procedure. 0-NOP-67.11, step 4.1, Initial Activities, requires review of Section 2.0, Precautions and Limitations. Precaution 2.1.1 states that the FME area requirements of MA-AA-101-1000, Foreign Material Exclusion Procedure, shall be followed during the performance of 0-NOP-67.11. MA-AA-101-1000, requires that MA-AA-101-1000-F08, FME Complex Equipment and Tooling Inspection Checklist, be completed when complex tooling, such as the SCOUT tool, are used in the performance of work. Contrary to the above, on September 16, 2021, during the St. Lucie Unit 2 fall 2021 refueling outage, MA-AA-101-1000-F08, FME Complex Equipment and Tooling Inspection Checklist, was not implemented during CEA coupling.

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

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

On October 27, 2022, the inspectors presented the integrated inspection results to Mr.

Carlos Santos, Operations Director, and other members of the licensee staff.

On August 12, 2022, the inspectors presented the Heat Sink inspection results to Don DeBoer, Site Vice President, and other members of the licensee staff.

On September 15, 2022, the inspectors presented the RP Routine Occupational and Public Radiation Safety Baseline inspection results to Dan DeBoer, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.07T Corrective Action 2035340 1A CCW heat exchanger SL1-26 inspection results 04/30/2015

Documents 2089161 Limited Hypochlorite Injection to ICW/CW 01/13/2016

2433921 2022 NRC Triennial Inspection. Level switch LS-21-SA has a 08/09/2022

detached/broken cable conduit

Corrective Action 2433926 Temperature indicator TI-21-13A has residual water which 08/09/2022

Documents indicate a degraded gauge condition.

Resulting from 2433929 FIS-21-9A Switch electrical junction box has 2 out 4 machine 08/09/2022

Inspection screws missing.

2433932 1A CCW HX shell side drains (blind flanges nuts and bolts) 08/09/2022

have surface corrosion.

2433934 Minor coating degradation on 2A Shutdown Cooling HX stand. 08/09/2022

2434043 CCW HX 1B End bell fasteners surface corrosion. 08/10/2022

Miscellaneous PSL 4th Interval ISI - SHUTDOWN COOLING HEAT 0

EXCHANGER 2A

SL-31 ICW Pipe Replacement Phases

Unit 1 ICW Chlorine samples 12 months 08/04/2022

Procedures 0-OSP-37.01 Emergency Cooling Water Canal - Periodic Test 15

0-PMM-14.01 COMPONENT COOLING WATER HEAT EXCHANGER 17

CLEAN / REPAIR

SPEC-C-004 Protective Coatings for Areas Outside the Reactor 15

Containment

Work Orders 40696099 08 U1 CCW HX 1A: CLEAN/INSPECT/ECT TEST 06/14/2021

40731176 08 U1 CCW HX 1B: Repair External Coat/grease Nuts & Stud 04/29/2021

71124.01 Corrective Action Condition Report

Documents Number

Resulting from 02436658

Inspection Condition Report

Number

2436836

71124.08 Corrective Action Condition Report

Documents Numbers

2422924,

Inspection Type Designation Description or Title Revision or

Procedure Date

2405720, and

2353235

Corrective Action Condition Report

Documents Number

Resulting from 02436905

Inspection Condition Report

Number 0243818

Procedures RP-AA-107-1002 Requirements for Radioactive Materials Stored Outdoors Rev. 10

Self-Assessments AR # 02249437 Part 37 Annual Review 12/13/2021

IMC 0609, APPENDIX M, TABLE 1

QUALITATIVE DECISION-MAKING ATTRIBUTES FOR NRC MANAGEMENT REVIEW

ST LUCIE UNIT 2 NCV 05000389/2022003-01: REACTOR SHUTDOWN REQUIRED BY

TECHNICAL SPECIFICATIONS DUE TO A MISALIGNED CONTROL ELEMENT ASSEMBLY

Decision Attribute Basis for Input to Decision - Provide

qualitative and/or quantitative information for

management review and decision making.

Defense-in-Depth RCS and containment barriers are not

expected to be impacted by the L-slot pin. The

pin is made of 410 stainless steel and is

approximately 0.4 inches long and 0.25 inches

in diameter. The loose part weighs

approximately 0.02 lbs.

A vendor analysis assessed potential impact

to the reactor internals, CEDM, reactor head

vent, pressurizer, RCS piping, RCPs, steam

generators, safety injection system, shutdown

cooling, CVCS, and primary chemistry. The

analysis concluded that the missing L-slot pin

condition has been bounded by other more

significant conditions that have been

previously evaluated at other CE and

Westinghouse plants.

A separate fuel vendor analysis assessed the

potential impact of the pin via entry through

the upper tie plate, between the fuel

assemblies, between the fuel assemblies and

shroud, through the center guide tube and

through the corner guide tube. The analysis

concluded that there was a small potential for

isolated fuel cladding damage that would be

detectable through primary chemistry

monitoring.

Issue Date: 01/10/19 T1-1 0609 App M

Safety Margin Impact on Rod Insertion

Appendix G to Volume 10 of NUREG/CR-

5500, Reliability Study: Combustion

Engineering Reactor Protection System,

1984-1998, utilized PRA control rod failure

criteria of seven or more CEAs failing to insert

into the core on demand. The vendor loose

parts evaluation determined that it was

possible for one CEA to be prevented from

rapidly inserting. The evaluation noted that

analyses have shown that the reactor can be

safely shutdown with the highest worth CEA

stuck in the fully withdrawn position.

To examine the quantitative sensitivity of

common mode failure of control rods to insert

due to mechanical interference, the NRC

SPAR model basic event RPS-VCF-FO-

MECH (CONTROL ROD ASSEMBLIES

FAILS TO INSERT) was increased by one

order of magnitude from 1.20E-06 to 1.20E-05

for a condition exposure time of one year.

The resulting internal events estimated

increase in core damage frequency was

5.66E-07/year. Because the loose part would

not be expected to impact enough control rods

to fail the SPAR model RPS rod insertion

function, the quantitative values in the

sensitivity would be expected to overestimate

this risk.

Impact on Fuel Cladding

The evaluation concluded there was a small

possibility for the L-slot pin to cause a debris

related fuel failure but monitoring of the core

chemistry would indicate if a debris-induced

failure had occurred and allow further

evaluation of the fuel assemblies and core.

Issue Date: 01/10/19 T1-2 0609 App M

Extent of Condition The single missing L-slot pin condition only

affected St. Lucie Unit 2. The other missing

SCOUT tool L-slot pin was retrieved from the

CEDM 27 motor following its removal from

Unit 2. The missing SCOUT tool roll pin was

determined to have been missing prior to its

use by St. Lucie during the Fall 2021 refueling

outage and, therefore, would not have had the

potential to be introduced into the Unit 2 RCS.

The vendor loose parts analysis concluded

that the missing L-slot pin was bounded by

parts that have been previously evaluated at

other Combustion Engineering and

Westinghouse-designed plants. The LPEV

concluded that the operation of St Lucie Unit 2

for one cycle of operation following the Spring

22 forced outage with the missing pins was

acceptable.

The fuel vendor analysis evaluated the risk

and potential impact on fuel assembly

performance based on the L-slot pin

characterization. The evaluation concluded

there was a small possibility for the L-slot pin

to cause a debris related fuel failure but

monitoring of the core chemistry would

indicate if a debris-induced failure had

occurred and allow further evaluation of the

fuel assemblies and core. The evaluation

additionally concluded that the CEAs on the

current cycle were inserted prior to the L-slot

pin being lost, therefore, safe shutdown of the

reactor was not a concern. The missing L-slot

pin also would also not prevent CEA insertion

or shutdown, would not sufficiently block

coolant flow to affect the thermal hydraulic

performance of a fuel assembly, and the L-slot

pin in its entirety or fragmented, has minimal

potential to promote corrosion.

Degree of Degradation St Lucie established procedure NF-AA-100-

1001, Failed Fuel Action Plan, in part to

identify specific action thresholds for the

evaluation of failed fuel using radio isotopic

data. The current Unit 2 radio isotopic

analyses results are well below established

thresholds indicative of a fuel clad defect.

Exposure Time Condition is assumed to last for maximum

SDP exposure time of 1-year. The missing L-

slot pin may have entered the RCS prior to

startup from a refuel outage on September 29,

21.

Issue Date: 01/10/19 T1-3 0609 App M

Recovery Actions The missing L-slot pin does have a

qualitatively assessed low potential of

increasing the consequences associated with

a core damage accident sequence through

damage to fuel cladding. No specific recovery

actions are credited but primary chemistry

monitoring would be expected to provide early

(pre-event) indication of clad damage that

could be assessed.

If the clad damage did not reveal itself until a

core damage sequence occurred, emergency

response procedures could provide indication

of cladding damage and an evaluation if

escalation of emergency response measures

was warranted.

Additional Qualitative Licensee corrective actions included replacing

Considerations CEDM 27, incorporating complex tool

inspection into the reactor services FME plan,

procedure revisions to address FME risks

during CEA coupling, and work order activities

to perform a foreign object search and

retrieval activity for the missing L-slot pin

during the next Unit 2 refueling outage.

Both vendor analysis of the condition

supported continued operation of the cycle

and recommended thorough inspections of

the top of fuels assemblies prior to core

offload and an inspection of the lower core

plate once the fuel assemblies have been

removed.

Result of management review (COLOR): GREEN

Drafted by: Shane Sandal, Senior Reactor Analyst

Reviewed by: David Dumbacher, Chief, Reactor Projects Branch 3

Issue Date: 01/10/19 T1-4 0609 App M