IR 05000250/2023004

Integrated Inspection Report 05000250/2023004 and 05000251/2023004

ML24043A263
Person / Time
Site:	Turkey Point
Issue date:	02/13/2024
From:	David Dumbacher, Jeffrey Hamman NRC/RGN-II/DRP/RPB3
To:	Coffey B Florida Power & Light Co
References
IR 2023004
Download: ML24043A263 (33)

Text

SUBJECT:

TURKEY POINT UNITS 3 & 4 - INTEGRATED INSPECTION REPORT 05000250/2023004 AND 05000251/2023004

Dear Bob Coffey:

On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Turkey Point Units 3 & 4. On February 1, 2024, the NRC inspectors discussed the results of this inspection with Michael Strope, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Eight findings of very low safety significance (Green) are documented in this report. Seven of these findings involved violations of NRC requirements, one was determined to be Severity Level IV. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

A licensee-identified violation which was determined to be Severity Level IV is documented in this report. We are treating this violation as an NCV, 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 Turkey Point Units 3 & 4.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 Turkey Point Units 3 & 4.

February 13, 2024 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, David E. Dumbacher, Chief Reactor Projects Branch 3 Division of Reactor Projects Docket Nos. 05000250 and 05000251 License Nos. DPR-31 and DPR-41 Enclosure:

As stated cc w/ encl: Distribution via LISTSERV Hamman, Jeffrey signing on behalf of Dumbacher, David on 02/13/24

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Turkey Point Units 3 & 4, 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. A licensee-identified non-cited violation is documented in report section: 7115

List of Findings and Violations

Inadequate 10 CFR 50.65(a)(4) Shutdown Risk Assessment Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000251/2023004-01 Open/Closed

[H.8] -

Procedure Adherence 71111.13 NRC inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), for the licensees failure to adequately assess and manage Unit 4 shutdown risk following a 4A-train loss of offsite power safeguards surveillance test failure.

Failure to Address Operability of the ECCS and Submit a 10 CFR 50.73 report Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Mitigating Systems Green Severity Level IV NCV 05000250/2023004-02 Open/Closed

[P.2] -

Evaluation 71111.15 The inspectors identified a Green finding and associated NCV of technical specification (TS)3.5.2, "ECCS Subsystems," for the licensees failure to consider the licensing and design bases requirements for the application of single failures when evaluating the emergency core cooling system (ECCS) subsystem operability; and a Severity Level (SL) IV NCV of 10 CFR 50.73(a) for the failure to issue a licensee event report.

Failure to Perform Required PMT on Diesel Engines for Unit 3 and Unit 4 Instrument Air Compressors 3CD and 4CD Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000250,05000251/2023004-03 Open/Closed

[H.5] - Work Management 71111.24 NRC inspectors identified a Green finding (FIN) when the licensee failed to perform testing as required by licensee procedures, on the diesel engines for Unit 3 and Unit 4 diesel-driven instrument air (IA) compressors 3CD and 4CD.

Failure to Test MOV Circuit Breaker in Accordance with Procedure Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000250/2023004-04 Open/Closed

[H.10] - Bases for Decisions 71111.24 NRC inspectors identified a Green finding and associated NCV of 10 CFR 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to comply with licensee procedures to perform required Post-Maintenance Testing (PMT) of circuit breaker 3N1403 which operates the steam supply to auxiliary feed water (AFW) pump motor operated valve (MOV) 3-1403. Specifically, following preventive maintenance (PM) activities on the motor starter for MOV 3-1403, the licensee did not perform PMT of circuit breaker 3N1403 as directed by safety-related procedure 0-PME-102.15, AFW N1403 and N1405 DC Reversing Starter Inspection.

Failure to Follow the 4A Safeguards Test Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000251/2023004-05 Open/Closed

[H.5] - Work Management 71111.24 A self-revealed Green finding and associated NCV of 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, was identified for the licensees failure to follow safety-related procedure 4-OSP-203.1, Train A Engineered Safeguards Integrated Test, which resulted in a loss of positive control of critical steps to verify installed fuses, which increased plant risk.

Failure to Implement Adequate Procedure Results in Unexpected Change in Radiological Conditions Requiring Locked High Radiation Area Controls Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000251/2023004-06 Open/Closed None (NPP)71152A A self-revealed Green finding and associated NCV of Turkey Point Unit 4 TS 6.12, High Radiation Area, was identified when chemical and volume control system (CVCS)demineralizer operations unexpectedly created high radiation area conditions greater than 1,000 mrem/hr at 30 cm, in an occupied area that was a posted Radiation Area. Specifically, licensee procedure 4-OP-047.3, CVCS - Demineralizer Operations, did not contain adequate instructions which would have prevented the unexpected change in radiological conditions experienced during the event.

Failure to Correct Bolting on Unit 3 RHR Recirculation Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000250/2023004-07 Open/Closed

[H.14] -

Conservative Bias 71152S NRC inspectors identified a Green finding and associated NCV of 10 CFR 50, Appendix B,

Criterion XVI, "Corrective Action," for the licensees failure to correct degraded bolting on MOVs 3-863 A & B (RHR heat exchanger discharge to the refueling water storage tank).

Unit 3 Reactor Trip While Performing Unit 4 RPS Testing Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000250/2023004-08 Open/Closed

[H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated NCV of 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, was identified for the licensees failure to ensure that Unit 4 reactor protection system (RPS) testing was executed as written in safety-related procedure 4-SMI-049.01B, Train B Reactor Protection System Logic Testing, which resulted in an unplanned Unit 3 reactor trip and subsequent transient.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000250/2023-003-00 Unit 3, Grid Disturbance in Switchyard Causes Automatic Reactor Trip 71153 Closed LER 05000250/2023-001-00 Unit 3, RCS Pressure Boundary Degraded 71153 Closed LER 05000250/2023-004-00 Unit 3, Unplanned Automatic Scram During RPS Testing 71153 Closed LER 05000250/2023-002-00 Unit 3, Manual Reactor Trip due to Steam Generator Feedwater Regulating Valve Positioner Failure 71153 Closed

PLANT STATUS

Unit 3 began the inspection period at full rated thermal power (RTP). On October 24, 2023, the unit scrammed due to unplanned reactor protection system testing (reference LER

===05000250/2023-004-00). The unit was restarted on October 25, 2023, and returned to full RTP on October 26, 2023, where it remained for the rest of the inspection period.

Unit 4 began the inspection period in the middle of a refueling outage. The unit began startup on November 11, 2023, and returned to full RTP on November 15, 2023, where it remained 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 onsite 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) The inspectors evaluated the adequacy of the overall preparations to protect risk significant systems from impending water intrusion from a severe rain event on November 15, 2023.

71111.05 - Fire Protection

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

(1) The inspectors evaluated the onsite fire brigade performance during an actual fire event on the second floor of the old maintenance building on December 14, 2023.

===71111.08P - Inservice Inspection Activities (PWR) The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from October 2, 2023 to October 6, 2023.

PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===

The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:

(1) Ultrasonic Testing (UT)

• 14"-FWA-2401-1922, Elbow to Pipe

• Penetration T4P32, Electrical Penetration to the Control Rod Drive Mechanisms (CRDM)

Penetrant Testing (PT)

• Penetration T4P32 Mockup, Electrical Penetration to the CRDM Magnetic Particle Testing (MT)

• 22-034, FW FAC Pipe Location 20 FW-1

==4R33 -018, FW FAC Location 20 FW-2

• 4R33-037, FW FAC Location 20 FW-3 FW-4 Visual Examination (VT)

• ==

Metallic Containment, Moisture Barrier

===80117-H-341-01, Double Acting Restraint

• Reactor Vessel Closure Head Bare Metal Visual Examination

• Penetration T4P32 Mockup, Electrical Penetration to the CRDM Welding Activities

• Shielded Metal Arc Welding (SMAW)o Feedwater Pipe FAC Location 20, FW-1/2/3/4 PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection Activities (IP Section 03.02)===

The inspectors verified that the license conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately:

(1)

• Reactor Vessel Closure Head Bare Metal Visual Examination PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)

The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:

(1)

• Boric Acid Walkdown - October 3, 2023

• Action Request (AR) 02461480

• AR 02461481

• AR 02461483

71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance

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

The licensee completed the annual requalification operating examinations and biennial written examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), "Requalification Requirements," of the NRC's "Operator's Licenses." The inspector performed an in-office review of the overall pass/fail results of the individual operating examinations, the crew simulator operating examinations, and the biennial written examinations in accordance with Inspection Procedure (IP) 71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance." These results were compared to the thresholds established in Section 3.03, "Requalification Examination Results," of IP 71111.11.

(1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exams completed by September 15, 2023 and the biennial written examinations completed by September 15, 2023.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)

(1) The inspectors observed and evaluated licensed operator performance in the Unit 3 and Unit 4 main control room during the following operational evolutions:

• Unit 4 train A loss of offsite power safeguards testing on October 5, 2023

• Unit 3 reactor start-up, Mode 2 and criticality on October 25, 2023

• Unit 4 main turbine spin test to heat up the generator to remove an electrical ground on October 30, 2023

• Unit 4 reactor startup through Mode 2 and criticality on November 11, 2023

• Unit 3 and 4 control rooms during an actual fire in the protected area on December 14, 2023

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

(1) The inspectors observed and evaluated Unit 3 start-up just-in-time training in the simulator on October 25, 2023.

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (4 Samples)

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

(1) Unit 3 online risk while at RTP, and Unit 4 yellow shutdown risk due to the RCS inventory less than 5% pressurizer cold calibration level and reactor time to boil at 14.45 minutes, on October 3, 2023

(2) Unit 3 online risk while at RTP, and Unit 4 yellow shutdown risk due to the safety-related 4B 4KV bus being unavailable for maintenance, on October 11, 2023

(3) Unit 3 online risk while at RTP, and Unit 4 yellow shutdown risk while in mode 4 due to RCS inventory less than 5% pressurizer cold calibration level, and the 4B boric acid transfer pump and containment radiation instrumentation R11/R12 unavailable

(4) Unit 4 shut down risk and Unit 3 online risk after an emergent issue when the 4B emergency diesel generator (EDG) became unavailable during safeguards testing, on October 5, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)

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

(1)operability assessment of residual heat removal (RHR) recirculation MOV's 3/4 MOV-863 A&B actuator to bonnet bolting that were torqued beyond assigned material yield strength

(2) AR 2450051, NRC concern for TS 3.5.2 and RHR Single Failure Criterion

(3) AR 2469221, 4-11-003 check valve exceeded allowed leakage for LLRT

(4) AR 2471181, 3C intake cooling water pump vibration data in alert range

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) Unit 3 and 4 RCP seal alarm modifications because of seal degradation

71111.20 - Refueling and Other Outage Activities

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

(1) The inspectors evaluated Unit 4 refueling outage, PTN 4-34, from September 29, 2023, to November 13, 2023.

71111.24 - Testing and Maintenance of Equipment Important to Risk

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

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

(1)4-OSP-013.3, Diesel Instrument Air Compressor Functional Test, after completing the 4CD diesel instrument air compressor 2-year preventive maintenance per work order (WO) 40828387-01 (2)4-387B, Excess Letdown Thermal Relief Check Valve IST/PMT, WO 4095335-01

(3) MOV 4-843B, SI Cold Leg Injection Valve diagnostic and PMT WO 40697551, on October 16, 2023

(4) MOV 4-381, RCP Seal Water Return and Excess Letdown Isolation, diagnostic and PMT WO 40829902, on October 23, 2023

(5) CV-4-2831 (TRAIN 2) - AFW to S/G A Control Valve - Diagnostic Testing and PMT WO 40829938, on October 8 - 23, 2023

(6) Unit 4 source range nuclear instrument N-4-32 calibration and test after replacing module, WO 40829829-01

(7) Replace impeller on the Unit 4B boric acid transfer pump and perform testing per WO

40679313, on October 18, 2023 Surveillance Testing (IP Section 03.01)

(1)4-OSP-203.1, Train A Engineered Safeguards Integrated Test, Section 7.3 - Loss of Offsite Power Coincident with Safety Injection, on October 6, 2023

71114.06 - Drill Evaluation

Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)

(1) The inspectors observed processes and procedures used for evaluation and notification of accident situations in the Technical Support Center and Emergency Operating Facility on November 28,

71124.06 - Radioactive Gaseous and Liquid Effluent Treatment

Walkdowns and Observations (IP Section 03.01) (3 Samples)

The inspectors evaluated the following radioactive effluent systems during walkdowns:

(1) Unit 3 spent fuel pool (SFP) gaseous effluent monitor no. RIC-3-6418

(2) Liquid effluent monitor no. RD-018

(3) Plant vent stack gaseous effluent monitors no. RAD-6304 A & B

Sampling and Analysis (IP Section 03.02) (3 Samples)

Inspectors evaluated the following effluent samples, sampling processes and compensatory samples:

(1) Release for gas decay tank no. F-GDT, October 11, 2023

(2) Release for liquid monitoring tank no. MT-A, October 15, 2023

(3) Release for liquid monitoring tank no. MT-B, October 16, 2023

Dose Calculations (IP Section 03.03) (3 Samples)

The inspectors evaluated the following dose calculations:

(1) Liquid effluents dose calculation for release permit no. L-2023-177

(2) Gaseous effluents dose calculation for release permit no. G-2023-117

(3) Gaseous effluents dose calculation for release permit no. G-2021-038

Abnormal Discharges (IP Section 03.04) (1 Sample)

The inspectors evaluated the following abnormal discharges:

(1) Unplanned liquid release from an open valve on 3A component cooling water heat exchanger vent to the north east storm drain, on August 7, 2022

71124.07 - Radiological Environmental Monitoring Program

Environmental Monitoring Equipment and Sampling (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated environmental monitoring equipment and observed collection of environmental samples.

Radiological Environmental Monitoring Program (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the implementation of the licensees radiological environmental monitoring program.

GPI Implementation (IP Section 03.03) (1 Sample)

(1) The inspectors evaluated the licensees implementation of the Groundwater Protection Initiative (GPI) program to identify incomplete or discontinued program elements.

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

MS06: Emergency AC Power Systems (IP Section 02.05)===

(1) Unit 3 (October 1, 2022 through September 30, 2023)

(2) Unit 4 (October 1, 2022 through September 30, 2023)

MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)

(1) Unit 3 (October 1, 2022 through September 30, 2023)

(2) Unit 4 (October 1, 2022 through September 30, 2023)

MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)

(1) Unit 3 (October 1, 2022 through September 30, 2023)

(2) Unit 4 (October 1, 2022 through September 30, 2023)

MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)

(1) Unit 3 (October 1, 2022 through September 30, 2023)

(2) Unit 4 (October 1, 2022 through September 30, 2023)

MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)

(1) Unit 3 (October 1, 2022 through September 30, 2023)

(2) Unit 4 (October 1, 2022 through September 30, 2023)

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) March 1, 2022 through September 30, 2023

71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1)completed corrective actions for NCV 05000250,05000251/2022011-01, MOV-4-863A&B, Weak Link Analysis Discrepancy, to verify they had been completed as scheduled, on October 23, 2023

(2) AR02470236, U4 A Engineering Safeguards Integrated Test Failure 4OSP-203.1, on October 21, 2023

(3) AR02467086, Unanticipated Dosimeter Alarm for Changing Conditions, on September 16, 2023

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

(1) The inspectors reviewed the licensee's corrective action program for potential adverse trends in the instrument air system that might be indicative of a more significant safety issue or concern. The inspectors identified a Green finding related to post maintenance testing that is documented in this report under Inspection Results Section 71111.24.

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)

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

(1) LER 05000250/2023-001-00, Unit 3, RCS Pressure Boundary Degraded (ADAMS Accession No. ML23165A158). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is Closed.

(2) LER 05000250/2023-002-00, Unit 3, Manual Reactor Trip due to Steam Generator Feedwater Regulating Valve Positioner Failure (ADAMS Accession No. ML23275A053). The inspectors reviewed the events surrounding this LER in inspection report 05000250/2023003 and 05000251/2023003, Section 71152A. No performance deficiencies were identified. This LER is Closed.

(3) LER 05000250/2023-003-00, Unit 3, Grid Disturbance in Switchyard Causes Automatic Reactor Trip (ADAMS Accession No. ML23324A275). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensee's ability to foresee and correct, and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified. This LER is Closed.

(4) LER 05000250/2023-004-00, Unit 3, Unplanned Automatic Scram During RPS Testing (ADAMS Accession No. ML23347A130). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is Closed.

INSPECTION RESULTS

Inadequate 10 CFR 50.65(a)(4) Shutdown Risk Assessment Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000251/2023004-01 Open/Closed

[H.8] -

Procedure Adherence 71111.13 NRC inspectors identified a Green finding and associated NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), for the licensees failure to adequately assess and manage Unit 4 shutdown risk following a 4A-train loss of offsite power safeguards surveillance test failure.

Description:

On October 5, 2023, the licensee performed Unit 4 surveillance procedure 4-OSP-203.1, Train A Engineered Safeguards Integrated Test, Section 7.1, Loss of Offsite Power Test, with Unit 4 in Mode 6, green shutdown risk, and Unit 3 at full RTP, green online risk.

During the surveillance test, on October 5, 2023 at 10:15, the 4A sequencer failed to strip the emergency bus immediately following the loss offsite power signal. The 4A EDG started but failed to load the necessary EDG auxiliary components, specifically the EDG cooling fans. This condition resulted in increasing EDG oil and cooling water temperatures causing smoke to develop in the 4A EDG room. At 10:28 on October 5, 2023, the 4A EDG was emergency stopped causing a lockout of the 4A EDG, which was subsequently declared inoperable and unavailable. The unavailability resulted in a change from green to yellow shutdown risk on Unit 4, while Unit 3 online risk remained green. The surveillance was declared unsatisfactory.

The licensees failure investigating process (FIP) team identified that test configuration issues were the cause of the test failure. The FIP team's review of the EDG's temperatures concluded the 4A EDG was stopped prior to any potential permanent damage. After corrective actions to restore the system configuration and reset the diesel lockout, the 4A EDG was declared available and Unit 4 shutdown risk transitioned back to green.

The licensee subsequently retracted their yellow shutdown risk assessment after concluding that 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, did not apply to this event. Control room logs show that this conclusion was based on the FIP teams determination that no permanent damage to the 4A EDG occurred. The licensee subsequently reassessed that Unit 4 had always remained in green shutdown risk by answering yes to Items 4 and 5 on Enclosure 12 of procedure 0-ADM-051, Outage Risk Assessment and Control. Item 4 states, Is the contingency plan for the restoration of the component well communicated and understood by all groups involved?

and Item 5 states, Is the contingency plan for the restoration of the component adequate and reasonably expected to be accomplished? The licensee stated they had established a verbal risk contingency plan sometime after the 4A EDG was declared unavailable (at time of 4A EDG lockout). However, this was not implemented until after the FIP team assessments (approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following the 4A EDG lockout).

The inspectors determined that 10 CFR 50.65(a)(4) requirements to assess and manage the increase in risk from maintenance activities, were applicable to this event's risk assessment. A review of licensee procedures confirmed that this event was within their ability to foresee. The inspectors found the licensee did not have a contingency plan in place prior to the event. In addition, procedure OM-AA-101-1000, Shutdown Risk Management, Section 4.7.4 states Shutdown Risk Contingency Plans may not be used to change a Safety Status Color to a less degraded color. Shutdown Risk Contingency Plans do not change the identified shutdown risk but provide a systematic means to respond to challenges to the availability of key system and components. The inspectors determined the 4A EDG was unavailable throughout the entire period of this event up to the time the 4A EDG lockout was reset. The licensee had incorrectly answered item 4 and item 5 on enclosure 12 to re-assess the event. Specifically, the licensee did not correctly implement the shutdown risk and contingency plan procedures, which led to incorrectly declaring the 4A EDG available thereby not addressing the requirements of 10 CFR 50.65 (a)(4) to assess and manage the increase in risk for this event.

Corrective Actions: The licensee entered this issue into the corrective action program to investigate the process used to retract the Unit 4 yellow shutdown risk.

Corrective Action References: ARs 02468709 and 022475170

Performance Assessment:

Performance Deficiency: The licensee's failure to adequately assess and manage Unit 4 shutdown risk in accordance with licensee procedures 0-ADM-051 and OM-AA-101-1000, following a 4A-train loss of offsite power safeguards surveillance test failure was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to correctly address maintenance risk prevents the site from taking necessary precautions during high-risk activities. They may not stop when faced with uncertain conditions and evaluate and manage risk before proceeding.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix K, Maintenance Risk Assessment and Risk Management SDP. A regional Senior Reactor Analyst (SRA) calculated incremental core damage probability (ICDP) for the maintenance using SAPHIRE 8 Version 8.2.9 and the Turkey Point Units 3 & 4 SPAR model version 8.81 dated 07/28/2023. The maintenance activity was modeled as the 4A EDG inoperable due to Test and Maintenance and a conservative exposure period of 1 day was assigned. Both the operating unit (Unit 3) and the outage unit (Unit 4) were considered. Unit 3 was in mode 1 and 100% power so an ICDP was calculated using the SPAR model. The dominate accident sequence was a weather related loss of offsite power, a common cause failure of the remaining EDGs, operators failing to declare extended loss of all AC power, failure of the Unit 4 turbine-driven auxiliary feedwater system, and failure to recover offsite power in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. ICDP was less than 1E-7 thus incremental large early release probability (ILERP) was also less than 1E-7.

Unit 4 was in mode 6 with the refueling cavity flooded up greater than 21 feet above the reactor vessel flange. The licensees qualitative shutdown risk assessment determined SFP cooling risk was green due to multiple trains of SFP cooling and RHR available prior to the maintenance evolution. The licensees qualitative shutdown risk assessment put the plant in a yellow risk configuration. Referring to IMC 0609 Appendix G, in this configuration due to the significant volume of coolant available and multiple trains of cooling available, the SRA determine the one shutdown initiating event of concern would be a mode 6 loss of inventory event. The Turkey Point SPAR model does contain shutdown event sequences so a quantitative assessment is possible. The SRA ran shutdown event trees for mode 6 and placed the 4A 4160-volt AC bus out of service to model the maintenance configuration and the 4A EDG failing to start. The dominant accident sequence was a loss of inventory event, with a subsequent loss of offsite power, failure of forced feed, and failure to enter low pressure recirculation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The ICDP would be less than 1 E-6 and the ILERP less than 1 E-7.

Because the performance deficiency was related to inadequate risk assessment, the inspectors were directed to IMC 0609 Appendix K Flowchart 1, Assessment of Risk Deficit. The inspectors determined the finding screened to very low safety significance (Green) since ICDP was less than 1E-6 and ILERP was less than 1E-7 for both units.

Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, the stations process, procedures, and work instructions to assess the availability of the 4A EDG was not followed and consequently resulted in declaring the 4A EDG available when it was not.

Enforcement:

Violation: 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," paragraph (a)(4), requires,in part, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Turkey Point procedures 0-ADM-051 and OM-AA-101-1000 implement those requirements for shutdown risk.

Contrary to the above, on October 5, 2023, during the Unit 4 outage after the 4A EDG failed safeguards testing, the licensee incorrectly assessed and managed the increase in risk after the cause of the emergent failure had been identified. Specifically, the licensee incorrectly assessed the Unit 4 shutdown risk as green, when Unit 4 was actually in a yellow shutdown risk configuration during the event.

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

Failure to Address Operability of the ECCS and Submit a 10 CFR 50.73 report Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Mitigating Systems Green Severity Level IV NCV 05000250/2023004-02 Open/Closed

[P.2] -

Evaluation 71111.15 The inspectors identified a Green finding and associated NCV TS 3.5.2, "ECCS Subsystems," for the licensees failure to consider the licensing and design bases requirements for the application of single failures when evaluating the ECCS subsystem operability; and an SL IV NCV of 10 CFR 50.73(a) for the failure to issue a licensee event report.

Description:

Manual valve 3-759A, located in the RHR system downstream of the A-RHR heat exchanger, failed to perform its safety function to close during testing on February 7, 2021, and November 10, 2021. The 3-759A & B valves are used to manually isolate the RHR discharge flow path from the reactor coolant system cold legs when aligned for long-term recirculation cooling if certain postulated single failures have occurred. The valve was repaired on March 9, 2023.

The inspectors determined that the ability for 3-759A & B to close serves a safety-related function when RHR is aligned for long-term core cooling. When aligning for long-term core cooling, the suction of one of the two RHR pumps is aligned to the containment sump, the normal low head safety injection (LHSI) path is isolated, and the discharge of the RHR/LHSI pumps is redirected to the suction of the high head safety injection (HHSI) pumps and the containment spray (CS) pumps; this is known as piggyback mode. The piggyback mode of operation is established, in part, by closing both RHR discharge to cold leg isolation MOVs, 3-744A and 3-744B, which are arranged in parallel, and opening the RHR alternate discharge isolation MOVs, 3-863A and 3-863B; this completely isolates the cold leg injection path and creates a suction path to the HHSI pumps and CS pumps.

The stations emergency operating procedures (EOPs) direct operators to manually close the 3-759 valves to isolate the cold legs if either of the MOV-3-744 valves fail to close. The EOP basis document states that both the 3-759 valves are closed, if either MOV-3-744A or MOV-3-744B fails to close, to prevent exceeding the RHR pump net positive suction head (NPSH)requirement when the system is in the piggyback configuration. If the NPSH requirement is exceeded, the RHR pump could cavitate and fail due to damage it sustains while operating in that condition. If the containment sump suction flowpath required by TS 3.5.2e. cannot provide adequate NPSH to its associated RHR pump, the ECCS safety function would be lost and would be considered inoperable.

The inspectors reviewed evaluations of the RHR pump NPSH in calculations CN-SEE-III-09-9, Turkey Point Safeguards Pump NPSH Analysis for EPU, revision 0, and CN-SEE-III-12-40, Turkey Point EPU - Evaluation of Revised Maximum Containment Spray and Residual Heat Removal Pump Performance Curves, revision 0. These calculations denote that, with MOV-3-744B and 3-759A open, the A RHR pump NPSH requirements would not be met.

The inspectors determined that the action statements for TS 3.5.2 were not satisfied.

Following inspector questioning regarding TS 3.5.2 operability, the licensee performed a past operability review (POR) using procedure EN-AA-203-1001, Operability Determinations/

Functionality Assessments, which was documented in AR 2450051. The licensees review did not declare any portion of the ECCS subsystems, inoperable. Updated Final Safety Analysis Report (UFSAR) Section 6.2.3, Single Failure Analysis, describes the licensing basis application of single failures associated with the ECCS. The inspectors determined that the licensees POR failed to consider single failures or demonstrate that the A RHR pump would have suitable NPSH with the MOV-3-744B valve (due to the postulated single failure) and 3-759A valve (due to the discovered degraded condition) in the open position. The TS bases clearly designate that the operability of the subsystems required by TS 3.5.2 depends on their availability in the event of a LOCA assuming any single active failure consideration. The degraded condition of 3-759A adversely affected the operability of the ECCS flow path and RHR pump capability when properly analyzed in accordance with the licensing and design bases.

As a result of the licensees misunderstanding of the licensing and design bases regarding single failures and incorrect conclusions in the POR, the licensee failed to make a 60-day report required by 10 CFR 50.73(a) for an event involving TS noncompliance or condition prohibited by TS.

Corrective Actions: The 3-759A valve was repaired and restored to operable on March 9, 2023. The issue regarding the incorrect operability assessment that led to inadequately assessing TS 3.5.2.e was entered into the corrective action program.

Corrective Action References: AR 2474058

Performance Assessment:

Performance Deficiency: The licensees failure to consider UFSAR Section 6.2.3 when evaluating the ECCS subsystem operability was a performance deficiency. Specifically, the licensee did not consider the impact that the failed 3-759A valve would have had on the flow path capability to perform its safety function during events that assume a postulated single active failure of the redundant B subsystem.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inability of the 3-759A valve to close affected the reliability and capability of the ECCS and containment spray systems to respond to initiating events.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The degraded condition represented a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time. Therefore, a detailed risk assessment was required.

A Regional SRA performed a detailed risk assessment using SAPHIRE 8 Version 8.2.6 and Turkey Point Unit 3 and 4 SPAR Model Version 8.80, dated May 26, 2022. The exposure time was set to one year (maximum allowed for an SDP) since 3-759A has been known to be stuck in the open position since at least 2021. The SPAR model only models 3-759 A/B as a normally open valve which could fail closed and MOV-3-744A/B as a normally closed valve that fails to open; these are the safety-related functions of the valves during the injection phase. The function of both valves to fail to close is not modeled directly. However, the high pressure recirculation function does require manual operator actions to realign the system.

Thus, the close function of these valves is indirectly modeled in the term HPI-XHE-XM-RECIRC as operators have to manually shut MOV-3-744A/B and if unsuccessful, manually close both 3-759A and 3-759B as directed by 3/4-EOP-ES 1.3. The SRA Adjusted HPI-XHM-XM-RECIRC (Operator Fails to Start/Control High Pressure Recirc - PWR) from 4E-3 to 4E-2 to account for a random failure of MOV-3-744A or MOV-3-744B or a loss of offsite power and failure of an EDG to run, resulting in a loss of the normal bus after the injection phase was commenced. The dominant accident sequence was a small break LOCA with a failure of operators to manually refill the Unit 3 refuel water storage tank, failure to cross-connect with Unit 4 high pressure injection, and failure to manually align for high pressure recirculation.

The change in core damage frequency was less the 1E-6 which corresponds to a finding of very low safety significance (Green)

Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. The licensee did not thoroughly evaluate the design basis and licensing basis impacts of the failed 3-759A valve commensurate with its safety significance during events involving the postulation of a single failure.

Enforcement:

Violation: TS 3.5.2, "ECCS Subsystems," requires, in part, The following Emergency Core Cooling System (ECCS) equipment and flow paths shall be OPERABLE." Item 3.5.2.e.

requires "Two flow paths capable of taking suction from the containment sump. The associated TS action statement required, that with one of the following components inoperable including an RHR suction flow path from the containment sump the licensee must restore the inoperable component to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or within the Risk Informed Completion Time Program or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Contrary to the above, from February 7, 2021, to March 9, 2023, because of the inoperable 3-759A valve, one of two ECCS flow paths capable of taking suction from the containment sump was inoperable and TS 3.5.2.e was not met, and the required TS action statement was not entered; the inoperable component was not restored within the required timeframe nor was the unit shutdown within the required timeframe. The suction flow path from the containment sump associated with the A-RHR pump was not capable of performing its safety function and the train was not declared inoperable nor was the action statement entered to mitigate the single failure concern.

Severity: The inspectors used the NRC Enforcement Policy to evaluate the traditional enforcement aspects of this issue. The NRC Enforcement Policy example in section 6.9.d.9 indicates the failure to make a 10 CFR 50.73(a) report was an SL IV NCV.

10 CFR 50.73(a), "Reportable Events," requires, in part, that the licensee shall submit an LER for any event of the type described in this paragraph within 60 days after the discovery of the event.

Contrary to the above, the licensee failed to submit an LER prior to 60 days from February 7, 2021, for an event involving TS noncompliance or condition prohibited by TS described in 10 CFR 50.73(a).

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

Failure to Perform Required PMT on Diesel Engines for Unit 3 and Unit 4 Instrument Air Compressors 3CD and 4CD Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000250,05000251/2023004-03 Open/Closed

[H.5] - Work Management 71111.24 NRC inspectors identified a Green finding (FIN) when the licensee failed to perform testing as required by licensee procedures, on the diesel engines for Unit 3 and Unit 4 diesel-driven IA compressors 3CD and 4CD.

Description:

The Unit 4 diesel-driven IA compressor (4CD) 2-year PM work activity was completed on July 21, 2023. The 2-year maintenance included significant work on the 4CD diesel engine and was subject to PMT requirements as described in maintenance procedure MA-AA-203-1000, Maintenance Testing, Attachment 6, Section 9. Operations surveillance procedure 3/4-OSP-013.3, Diesel Instrument Air Compressor Functional Test, implements the testing requirements for the 3CD and 4CD diesel-driven IA compressors, while Section 4.3, 4CD IA Diesel Compressor Local Start Test for PMT, is the required PMT to perform following the 2-year PM.

The inspectors noted that, following the 4CD maintenance on July 21, 2023, Operations completed Section 4.2, 4CD Low Pressure Auto Start Test, rather than Section 4.3, as required, and declared the diesel engine for the 4CD IA compressor back in service. Section 4.2 is a monthly test used to verify the low-pressure compressor auto start function, and is not a PMT for the diesel engine, which was the focus of the 2-year PM. Section 4.2 did not confirm the reliability of the diesel engine after completion of significant work. The diesel engine was not tested as required per the licensees PMT procedure. Additionally, the inspectors reviewed the work order instructions and identified that the required PMT was not referenced, and operations did not verify with work control the correct PMT to perform.

The licensee entered this issue into the corrective action program and determined that the incorrect test was performed. A review of the Unit 3 PMT for the 2-year IA compressor 3CD PM, completed on August 04, 2023, determined that it also used the incorrect test as the PMT. This WO also did not specify a PMT to use.

Corrective Actions: The licensee performed PMT Section 4.3 of procedure 3/4 -OSP-013.3 on the Unit 3 and Unit 4 diesel engines for the 3CD and 4CD IA compressors. A PMT task was also added to the model WO for the 2-year PM to include Operations testing per Section 4.3 of 3/4-OSP-013.3 as a PMT.

Corrective Action References: AR 2470068

Performance Assessment:

Performance Deficiency: The licensees failure to perform PMT on the diesel engines for IA compressors 3CD and 4CD as required by procedure 3/4 -OSP-013.3, 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 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 failure to perform the correct PMT and ensure the availability and reliability of the diesel engines for the 3CD and 4CD IA compressors adversely impacted the cornerstone objective to limit the likelihood of events that upset plant stability.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because it did not cause a reactor trip coincident with the loss of mitigation equipment relied upon to transition the plant from the onset of a reactor trip to a stable shutdown condition.

Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, when it was identified there was no PMT procedure referenced on the work orders, Operations and Work controls did not coordinate to ensure the correct PMT was performed.

Enforcement:

Inspectors did not identify a violation of regulatory requirements associated with this finding.

Failure to Test MOV Circuit Breaker in Accordance with Procedure Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000250/2023004-04 Open/Closed

[H.10] - Bases for Decisions 71111.24 NRC inspectors identified a Green finding and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to comply with licensee procedures to perform required PMT of circuit breaker 3N1403 which operates the steam supply to AFW pump MOV 3-1403. Specifically, following PM activities on the motor starter for MOV 3-1403, the licensee did not perform PMT of circuit breaker 3N1403 as directed by safety-related procedure 0-PME-102.15, AFW N1403 and N1405 DC Reversing Starter Inspection.

Description:

The inspectors observed the PM activities associated with the motor starter for MOV 3-1403, controlled by WO 40807629. The motor stater is required to be qualified to operate in accordance with 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants. Circuit breaker 3N1403 is important to the qualification. Testing of the motor starter was controlled by safety-related procedure 0-PME-102.15. Section 4.3 step 1 instructs the technician to perform the associated breaker testing using 0-PME-007.04, Molded Case Circuit Breaker Testing, on breaker under test. These procedures instruct the technicians to disassemble the motor starter and perform bench testing of the various parts individually at the maintenance facility.

The technicians did not follow the procedure to test the circuit breaker for the motor starter. Procedure 0-PME-007.04 was started and not completed. No testing was conducted.

The motor starter was reassembled and put into service without satisfying the appropriate quantitative acceptance criteria for determining that components important to 10 CFR 50.49 were satisfactorily accomplished.

Corrective Actions: The licensee placed the issue into the corrective action program and determined that the circuit breaker was successfully tested in 2018 using the same procedures under WO 40555280.

Corrective Action References: ARs 02467841 and 02470297

Performance Assessment:

Performance Deficiency: The failure to perform required performance verification testing in accordance with procedure 0-PME-102.15 on environmentally qualified circuit breaker 3N1403 was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to follow procedures to ensure the performance of safety-related circuit breakers that must operate in harsh environments affected the reliability and capability when required to respond to initiating events.

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, Item A.1, the finding screened to Green because it was a deficiency affecting the design or qualification of a mitigating SSC that maintained its operability or PRA functionality. The circuit breaker was successfully tested in 2018 using the same procedures under WO 40555280.

Cross-Cutting Aspect: H.10 - Bases for Decisions: Leaders ensure that the bases for operational and organizational decisions are communicated in a timely manner. Site management and maintenance personnel did not ensure that the basis for the decision to not test the circuit breaker was communicated in a timely manner so that the circuit breakers safety functions and qualification could be assessed based on that decision.

Enforcement:

Violation: 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, activities affecting quality shall be accomplished in accordance with instructions, procedures, or drawings.

Contrary to the above, on September 1, 2023, the licensee failed to accomplish safety-related procedure 0-PME-102.15 to perform required testing of circuit breaker 3N1403 to satisfy the appropriate quantitative acceptance criteria for determining that 3N1403 could perform its function in accordance with the 10 CFR 50.49 qualification.

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

Failure to Follow the 4A Safeguards Test Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000251/2023004-05 Open/Closed

[H.5] - Work Management 71111.24 A self-revealed Green finding and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to follow safety-related procedure 4-OSP-203.1, Train A Engineered Safeguards Integrated Test, which resulted in a loss of positive control of critical steps to verify installed fuses, which increased plant risk.

Description:

On October 5, 2023, the licensee implemented Unit 4 procedure 4-OSP-203.1 of which Section 7.1, Loss of Offsite Power Test, required that the alignment of the electrical systems to be verified. The test director hold point directly after step 7.1.6 required verification that 4kV breakers on the test train have been inspected to ensure proper jumper and T-bar alignment AND that all fuses are installed. This hold point was not performed in the sequence as it was written. As a result, the 4A emergency bus was not stripped. The 4A EDG was locked out and declared inoperable and unavailable. Unit 4 transitioned to yellow shutdown risk and the test was declared unsatisfactory.

The fuses, referred to in the hold point, for the circulating water pump breakers 4AA16 and 4AA18 were removed on the night of September 29, 2023, prior to the test for unrelated maintenance. This prevented the pumps from being stripped from the bus. The inspectors found that the test director hold point (step 7.1.6) to verify fuses installed was performed during the day of September 29, 2023. However, the test director hold point was not signed off until October 5, 2023, prior to starting the test. This led to a loss of positive control of critical procedure steps which was contrary to the stations continuous use procedure requirements (AD-AA-100-1006, Procedure and Work Instruction Use and Adherence).

Corrective Actions: The licensee entered this issue into the corrective action program to investigate the reason(s) the surveillance was executed with the breaker fuses not installed.

Corrective Action References: ARs 02468709 and 02474202

Performance Assessment:

Performance Deficiency: The failure to follow safety-related procedure 4-OSP-203.1 as written and lose positive control of critical steps to verify installed fuses, was a performance deficiency. Specifically, the failure to follow the procedure resulted in 4A EDG inoperability and unavailability and increased Unit 4 shutdown risk.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to follow the procedure resulted in making the 4A EDG and emergency bus inoperable and unavailable and increased the Unit 4 shutdown risk.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Safety SDP. The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 3

- Mitigating Systems Screening Questions, Item A, Mitigating Structure System Component (SSC) and PRA Functionality. The finding screened as Green because it did not represent an actual loss of safety function of one or more non-TS Trains of equipment during shutdown designated as risk-significant (e.g., 10CFR50.65), for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, site management did not ensure that: the organization implemented a process to plan, control, and implement the safeguards test that included positive control of the safeguards equipment at all times; the identification and management of the risk with contingency plans; and the coordination with the different groups that had access and cause to change the configuration of the safety bus switchgear.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion V, Instructions. Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with instructions, procedures, or drawings.

Contrary to the above, on October 5, 2023, the site failed to accomplish safeguards testing in accordance with instructions written in 4-OSP-203.1. Specifically, the failure to follow the procedure resulted in the lockout of the 4A EDG adversely affecting the shutdown risk to the plant.

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 Adequate Procedure Results in Unexpected Change in Radiological Conditions Requiring Locked High Radiation Area Controls Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000251/2023004-06 Open/Closed None (NPP)71152A A self-revealed Green finding and associated NCV of Turkey Point Unit 4 TS 6.12, High Radiation Area, was identified when CVCS demineralizer operations unexpectedly created high radiation area conditions greater than 1,000 mrem/hr at 30 cm, in an occupied area that was a posted Radiation Area. Specifically, licensee procedure 4-OP-047.3, CVCS - Demineralizer Operations, did not contain adequate instructions which would have prevented the unexpected change in radiological conditions experienced during the event.

Description:

On September 16, 2023, Operations was performing an evolution to re-introduce primary water flow through Unit 4 CVCS demineralizer 4E, which had just been loaded with fresh resin. Shortly after the evolution began, dose rates in the Unit 4 demineralizer valve gallery, initially posted as a radiation area, rose significantly, which resulted in an auxiliary operator in the demineralizer valve gallery receiving an unexpected dose rate alarm of 1090 mrem/hr with a setpoint of 300 mrem/hr.

Dose rates in the area, which were initially less than 100 mrem/hr general area, had increased to greater than 1,000 mrem/hr general area. A follow-up radiological survey indicated dose rates up to 2500 mrem/hr at 30cm. Upon receiving the dose rate alarm, the operator exited the area and reported to radiation protection (RP). RP technicians subsequently secured the area, which included posting it as a locked high radiation area (LHRA).

The licensee determined that the most probable cause was the collapse of an air void in the 4E demineralizer, which had just been filled with fresh resin, causing a migration of radioactive resin from the 4B demineralizer into the letdown line. This caused an unexpected increase in dose rates in the Unit 4 demineralizer valve gallery. The inspectors noted that the licensee had a demineralizer system that was unique from other Westinghouse designs, and that the potential off-normal operating scenarios were not well understood. A contributing factor was that licensee procedure 4-OP-047.3 did not include a requirement to isolate the demineralizers containing radioactive resin from the demineralizer being filled, which would have prevented the event.

Corrective Actions: The licensee has taken action to revise procedure 4-OP-047.3 with steps to isolate all other demineralizers when rinsing in a freshly loaded demineralizer.

Corrective Action References: The licensee placed this event into their corrective action program under AR 02467086.

Performance Assessment:

Performance Deficiency: The failure to implement adequate procedures for CVCS, including letdown and purification, as required by the Quality Assurance Topical Report under TS 6.8, Procedures and Programs, was a performance deficiency and was reasonably within the licensees ability to foresee and correct.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the performance deficiency is more than minor because it involves a high radiation area with greater than 1,000 mrem/hr at 30 cm, which was improperly posted and occupied by personnel.

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 planning, did not result in an overexposure beyond regulatory limits, there was no substantial potential for overexposure, and the ability to assess dose was not compromised.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned because the inspectors determined that the performance deficiency was rooted in longstanding vulnerabilities in procedures and demineralizer system design and therefore not reflective of current performance.

Enforcement:

Violation: Turkey Point Unit 4 TS 6.12, High Radiation Area, Section 6.12.2 states, in part, that areas accessible to personnel with radiation levels greater than 1,000 mrem/hr at 30 cm shall be posted, barricaded and provided with locked doors. Contrary to the above requirements, on September 16, 2023, operations in the Turkey Point Unit 4 demineralizer valve gallery caused dose rates in the area to increase above 1,000 mrem/hr at 30 cm while the area was not posted, barricaded, and provided with locked doors.

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

Failure to Correct Bolting on Unit 3 RHR Recirculation Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000250/2023004-07 Open/Closed

[H.14] -

Conservative Bias 71152S NRC inspectors identified a Green finding and associated NCV of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for the licensees failure to correct degraded bolting on MOVs 3-863 A & B (RHR heat exchanger discharge to the refueling water storage tank).

Description:

During an inspection of valve performance, the inspectors assessed the RHR valves that are required to perform safety functions during the recirculation mode of a loss of coolant accident. The licensee was previously issued a Green NCV (05000250,05000251/2022011-01) on December 20, 2022, because the valve operators bolts were preloaded with a torque value that exceeded the minimum yield strength (30 ksi) of the bolts by more than 200% at all temperatures and approached the 75 ksi tensile strength. The licensee entered the issue into the corrective action program and committed to correct the issue at the first available opportunity with the valves scheduled for repair by June 15, 2023.

The licensee had the opportunity to repair the MOV bolts during the U3 spring 2023 refueling outage but did not repair the valves. The work was canceled and rescheduled for December 6, 2024. The licensee justified this based on a prompt operability determination (POD) that determined the bolts were operable but degraded. The POD implied that the bolts had been repeatedly overstressed over years of maintenance activities and had not yet failed. It concluded that, It is probable that the bolting provided has a higher yield strength than 30 ksi as no substantial damage or stretching has been reported. These conclusions were made despite the clear 30 ksi bolt markings that were specified by the American Society of Mechanical Engineers (ASME) code requirements. Using this assumption did not provide reasonable assurance that the valves would perform their safety functions when required.

Corrective Actions: The licensee entered the issue into the corrective action program and committed to correct the valve operator during the next Unit 3 outage.

Corrective Action References: AR 2478058

Performance Assessment:

Performance Deficiency: The failure to correct a condition adverse to quality, the previously identified bolt deficiency on MOVs 3-863 A & B, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct bolted joint deficiencies to established acceptance criteria prolonged the adverse effect on availability and reliability of MOVs that respond to initiating events.

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, Item A.5 was most appropriate because the degraded condition represented a loss of a PRA system and/or function as defined in the Plant Risk Information Book. A regional SRA performed a risk analysis.

The detailed risk assessment used the guidance in Appendix A and using SAPHIRE 8 Version 8.2.6 and the Turkey Point SPAR model Revision 8.80 dated 5/26/2022 to model the condition. The SRA assumed an exposure time of 1-year (max allowed by SDP) and modeled the condition by adjusting the common cause failure terms for the failure to run of the motor driven RHR pumps. Since MOV 863A&B are 8 valves, Internal flooding would have to be considered as well but it is isolable. The SRA conservatively chose a one order of magnitude adjustment to the nominal CCF terms to bound the condition. The dominant accident sequence was a medium break loss of coolant accident with a failure of low pressure and high pressure recirculation. The increase in core damage probability was less the 1E-8.

Therefore, the finding is characterized as very low safety significance (GREEN)

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee did not use decision making practices that emphasized prudent implementation of corrective actions over those decisions that are simply allowable and failed to implement corrective actions that were determined to be safe, rather than those that were unsafe.

Enforcement:

Violation: 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," states, in part, that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected."

Contrary to the above, since June 15, 2023, the licensee failed to ensure that conditions adverse to quality, such as deficiencies, deviations, defective material and nonconformances were promptly corrected. Specifically, degraded RHR bolts that were deficiently torqued to values that deviated from material requirements were not corrected when the licensee had the opportunity to do so.

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

Unit 3 Reactor Trip While Performing Unit 4 RPS Testing Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000250/2023004-08 Open/Closed

[H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to ensure that Unit 4 RPS testing was executed as written in safety-related procedure 4-SMI-049.01B, Train B Reactor Protection System Logic Testing, which resulted in an unplanned Unit 3 reactor trip and subsequent transient.

Description:

On October 24, 2023, while Unit 4 was in Mode 3, maintenance technicians were testing the operability of the Unit 4 RPS per procedure 4-SMI-049.01B, Train B Reactor Protection System Logic Testing. Section 7.1.17 of this procedure contains instructions to test the reactor trip breakers at the clearly marked Unit 4 rack 41 and to ensure the breakers open after depressing selector pushbutton S4 on Unit 4 rack 41. Maintenance technicians mistakenly entered the Unit 3 rack 41 cabinet and manipulated Unit 3 RPS instead of Unit 4 RPS, causing and unplanned Unit 3 reactor trip from 100% RTP. The inspectors noted that all safety-related equipment functioned as designed following the Unit 3 reactor trip. Unit 3 was restarted on October 25, 2023, and reached full RTP on October 26, 2023.

Corrective Actions: Immediate corrective actions included a maintenance work stand down.

The technicians involved were remediated and tested on the core procedural human performance tools before returning to work. The licensee initiated a root cause team to investigate the issues that led to the Unit trip and to develop corrective actions.

Corrective Action References: AR 2470610

Performance Assessment:

Performance Deficiency: The licensees failure to ensure that Unit 4 RPS testing was executed as written in safety-related procedure 4-SMI-049.01B was a performance deficiency. Specifically, maintenance technicians entered the incorrect RPS cabinet (Unit 3 rack 41) and manipulated Unit 3 RPS components instead of manipulating Unit 4 RPS on Unit 4 rack 41 as instructed by 4-SMI-049.01B.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance 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 failure to perform testing in the Unit 4 RPS cabinet as instructed by procedure caused an event that upset plant stability and challenged critical safety functions during a Unit 4 refueling outage and Unit 3 at full power operation.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because it did not cause a reactor trip coincident with the loss of mitigation equipment relied upon to transition the plant from the onset of a reactor trip to a stable shutdown condition.

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. Specifically, management did not plan for the possibility of mistakes while executing procedures, latent issues with quality assurance indoctrination and inherent risk of working in the RPS, even while expecting successful outcomes from the assumption that personnel meet site performance standards. Management did not ensure that error reduction human performance tools were implemented.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with instructions, procedures, or drawings.

Contrary to the above, on October 24, 2023, the site failed to accomplish activities affecting quality in accordance with instructions, procedures, or drawings. Specifically, during execution of safety-related procedure 4-SMI-049.01B, Section 7.1.17, the licensee failed to complete the procedure steps as instructed when maintenance technicians mistakenly manipulated control switches on the Unit 3 RPS causing an unplanned Unit 3 reactor trip.

Disposition of this violation closes LER 05000250/2023-004-00, Unplanned Automatic Scram During RPS Testing.

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

Licensee-Identified Non-Cited Violation 71153 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: 10 CFR 50.72(b)(ii)(A) requires, in part, that the licensee shall notify the NRC Operations Center via the Emergency Notification System within eight hours of the occurrence of any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

Contrary to the above, the licensee failed to notify the NRC Operations Center within eight hours of the discovery of a degraded RCS pressure boundary. On April 8, 2023, during a refueling outage walkdown, the licensee identified a brown-colored dry boric acid deposit in the vicinity of seal table guide tube H-6. On April 15, 2023, a nondestructive examination revealed a degraded weld (crack) on the seal table guide tube, which did not meet American Society of Mechanical Engineers (ASME)Section XI code requirements. The licensee determined on April 17, 2023, two days after discovery, that the condition was reportable per 10 CFR50.72(b)(3)(ii)(A) due to a degraded principal safety barrier and should have been reported to the NRC Operations Center on April 15, 2023. On April 18, 2023, at 03:56, the licensee reported the degraded condition to the NRC Operations Center (EN#56474).

Severity: SL IV. Traditional enforcement is applicable to the violation because it could potentially impact the regulatory process. NRC Enforcement Policy Section 6.9.d.9 lists a failure to make a report required by 10 CFR 50.72 as an SL IV violation.

The disposition of this violation closes LER 05000250/2023-001-00, "Unit 3, RCS Pressure Boundary Degraded."

Corrective Action References: AR

EXIT MEETINGS AND DEBRIEFS

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

• On October 5, 2023, the inspectors presented the in-service inspection results to Chad Mynhier, Engineering Director, and other members of the licensee staff.

• On November 3, 2023, the inspectors presented the public radiation safety inspection results to Kyle Barry, Acting Site Vice President, and other members of the licensee staff.

• On December 1, 2023, the inspectors presented the occupational radiation safety inspection results, including disposition of a Green NCV for the LHRA event, to Michael Strope, Site Vice President, and other members of the licensee staff.

• On February 1, 2024, the inspectors presented the integrated inspection results to Michael Strope, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.01

Procedures

OP-AA-102-1002-

10000

Water Intrusion

Rev 2

71111.05

Procedures

0-ONOP-016.8

Response to Fire Smoke Detection System Alarm

71111.15

Corrective Action

Documents

Resulting from

Inspection

AR2441048

22 POV INSP MOV-4-8638 Weak Link Analysis

Discrepancy

11/30/2022

71124.06

Corrective Action

Documents

ARs 02461455,

2406873,

2471038,

2470467,

2433824, and

2438816

Corrective Action

Documents

Resulting from

Inspection

2471450

11/02/2023

22 Turkey Point Annual Radiological Effluent Release

Report

03/01/2023

22 Turkey Point Annual Radiological Environmental

Operating Report

03/12/2023

71124.07

Miscellaneous

Turkey Point Offsite Dose Calculation Manual (ODCM)

Rev. 29

Corrective Action

Documents

AR 02467086

Unanticipated dosimeter alarm for changing conditions

09/16/2023

Drawings

5614-M-3047

Chemical and Volume Control System, Charging and

Letdown

Rev 24

4-NOP-061.07E

CVCS Demineralizer 4E Resin Transfers

Rev 14

Procedures

4-OP-047.3

CVCS - Demineralizer Operations

Rev 26

PTN-M-2023-

0917-9

U4 Demin Valve Gallery Verification

09/17/2023

71152A

Radiation

Surveys

PTN-M-

230907-2

U-3 and U-4 Demin Valve Gallery Monthly

09/07/2023

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

PTN-M-

230916-1

U4 Demin Valve Gallery

09/16/2023

Radiation Work

Permits (RWPs)

23-0002

Operations Department Routine Activities

Rev 0