IR 05000327/2024002

Integrated Inspection Report 05000327/2024002 and 05000328/2024002

ML24212A056
Person / Time
Site:	Sequoyah
Issue date:	07/31/2024
From:	Louis Mckown Division Reactor Projects II
To:	Jim Barstow Tennessee Valley Authority
References
IR 2024002
Download: ML24212A056 (1)

Text

SUBJECT:

SEQUOYAH, UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000327/2024002 AND 05000328/2024002

Dear Jim Barstow:

On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Sequoyah, Units 1 and 2. On July 22, 2024, the NRC inspectors discussed the results of this inspection with Tom Marshall and other members of your staff. The results of this inspection are documented in the enclosed report.

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

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

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Sequoyah, Units 1 and 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 Sequoyah, Units 1 and 2.July 31, 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, Louis J. McKown, II, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket Nos. 05000327 and 05000328 License Nos. DPR-77 and DPR-79

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000327 and 05000328

License Numbers: DPR-77 and DPR-79

Report Numbers: 05000327/2024002 and 05000328/2024002

Licensee: Tennessee Valley Authority

Facility: Sequoyah, Units 1 and 2

Location: Soddy Daisy, TN 37379

Inspection Dates: April 01, 2024 to June 30, 2024

Inspectors: J. Bell, Senior Health Physicist D. Hardage, Senior Resident Inspector M. Magyar, Allegations/Enforcement Specialist P. Meier, Senior Resident Inspector A. Nielsen, Senior Health Physicist A. Price, Resident Inspector J. Vasquez, Reactor Inspector

Approved By: Louis J. McKown, II, Chief Reactor Projects Branch 5 Division of Reactor Projects

Enclosure

SUMMARY

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

Failure to Survey for Airborne Radioactivity Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.8] - 71124.01 Radiation Safety NCV 05000327,05000328/2024002-01 Procedure Open/Closed Adherence An NRC-identified Green NCV of TS 5.4.1, Procedures, was identified for the licensee's failure to implement an air-sampling procedure during radiological protection job coverage of work in the Unit 1 containment keyway. Specifically, the licensee failed to obtain an air sample while performing work in an area with removable contamination levels greater than 100,000 disintegrations per minute (dpm) per 100 cm2.

Inadvertent isolation of shutdown cooling Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.7] - 71153 NCV 05000327/2024002-02 Documentation Open/Closed A self-revealed Green finding and associated non-cited violation of Technical Specification (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement written procedures for activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Specifically, the licensee failed to establish and maintain adequate procedures for reactor coolant system (RCS) drain and fill operations. As a result, main control room operators unintentionally isolated shutdown cooling (SDC) flow when responding to a loss of communication with a remote valve operator during reactor cavity drain down.

Additional Tracking Items

None.

PLANT STATUS

Unit 1 began the inspection period shutdown for refueling outage 1R26. Reactor startup was performed on April 29, 2024, and the unit returned to rated thermal power (RTP) on May 3, 2024. On June 12, 2024, the unit was down-powered to 75 percent RTP to support switching in the 500kV switchyard. The unit returned to 100 percent RTP on June 13, 2024 and remained at or near RTP for the remainder of the inspection period.

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 due a tornado watch on April 2, 2024.

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 residual heat removal system (RHR) aligned for reactor coolant system (RCS)cooldown with RCS level below the flange at mid loop for nozzle dam removal on April 14, 2024

(2) Unit 1 RHR aligned for low head safety injection in Mode 1 following refueling outage

1R26 on May 8, 2024 (3)

2A emergency diesel generator (EDG) following a 24-hour surveillance run and load rejection test on May 30, 2024

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

(1) Unit 1 reactor building annulus on April 22, 2024

(2) Unit 1 lower containment on April 26, 2024

(3) Unit 1 upper containment on April 26, 2024

(4) EDG building elevation 722 on June 2, 2024

(5) Control building elevation 685 on June 16, 2024

===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 in Unit 1 during refueling outage 1R26 April 1 to April 5.

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 Examination

• N11, Inlet Nozzle, ASME Class 1

• W02-03, Bottom Head to Lower Middle Shell Circ. Weld, ASME Class 1

Visual Examination (VT)

• Bare metal visual of the Reactor Vessel Closure Head, N-729-6

• Upper Core Barrel Girth Weld, Outer Surface, MRP-227

• Baffle Edge Bolts, MRP-227

PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection

Activities (IP Section 03.02) (1 Sample)

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

(1)

• Bare metal visual of the Reactor Vessel Closure Head, N-729-6

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 - April 1, 2024

• CR 1893772

• CR 1921457

• CR 1921283

PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04) (1 Sample)

The inspectors verified that the licensee is monitoring the steam generator tube integrity appropriately through a review of the following examinations:

(1)

• Eddy Current Examination (ECT)o Steam Generator 1 - ECT for tubes R87C73, R98C66, R75C73 o Steam Generator 4 - ECT for tubes R29C61

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 main control room during RCS drain down to mid loop conditions in Mode 6 for steam generator nozzle dam removal on April 14, 2024.

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

(1) The inspectors observed and evaluated cycle 24-2 simulator exam 1 involving a loss of offsite power and steam-line break on May 23, 2024.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (1 Sample)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

(1) Unit 1 RHR pump discharge relief valve, 1-VLV-63-626, due to its failure to reseat after opening and resulting loss of RCS inventory and inoperability of the A train RHR on April 24, 2024

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 1 orange shutdown risk while RCS level was below the reactor flange at mid loop for nozzle dam removal on April 14, 2024, through April 15, 2024

(2) Unit 1 transition from mode 5 to mode 1 during the week of April 27, 2024, through May 1, 2024

(3) Unit 1 and unit 2 risk associated with the scheduled maintenance on 2B containment spray pump, 1B-B shutdown board transformer, and 2A safety injection pump during the week of May 19, 2024, through May 25, 2024

(4) Unit 1 and unit 2 risk associated with a unit 1 down power to 75 percent RTP for corrective maintenance in the switchyard, planned 1A EDG run, and planned maintenance on the 1C condensate booster pump during the week of June 10, 2024, through June 14, 2024

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) 1A RHR loop operability for mode 6 with the water level less than 23 feet above the top of the reactor vessel flange on April 10, 2024

(2) Loose coatings and concrete degradation in unit 1 lower containment due to refueling cavity leakage on April 16, 2024

(3) Low temperature overpressure protection (LTOP) system operability with the RCS depressurized for work on one of the power operated relief valves on April 19, 2024

(4) 1A containment spray heat exchanger inspection discovered with 127 damaged tubes due to exceeding shell side differential pressure on April 24, 2024

(5) Abnormally high oil level in the unit 2 turbine driven auxiliary pump lubricating oil sight glass due to water intrusion on May 10, 2024

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) Temporary design change package SQN-2-2024-063-01, pre-lube of 2A safety injection pump prior to normal starts, on June 12, 2024

71111.20 - Refueling and Other Outage Activities

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

(1) The inspectors evaluated refueling outage 1R26 activities from April 1, 2024, through April 30, 2024

71111.24 - Testing and Maintenance of Equipment Important to Risk

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

Post-Maintenance Testing (PMT) (IP Section 03.01) (5 Samples)

(1) Replace 1A-A RHR pump on April 11, 2024 (WO 117809230)

(2) Replace unit 1 RHR discharge to cold leg 2 and cold leg 3 relief valve, 1-VLV-63-626, on April 25, 2024 (WO 124448449)

(3) Unit 1 loop 4 crossover drain valve, 1-VLV-068-0557, boron cleaning and packing torque on April 29, 2024 (WO 124041012)

(4) Install newly refurbished and certified unit 1 pressurizer safety valve, VLV-68-563, on April 29, 2024 (WO 123267083)

(5) Replace unit 1 power operated relief valve, PCV-68-334B, on April 29, 2024 (WO

124382228) Surveillance Testing (IP Section 03.01)

(1) Unit 1 safety injection hot leg secondary check valve integrity test (1-SI-SXV-063-202.0) on April 28, 2024

(2) 2A-A EDG 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run and load rejection test (2-SI-OPS-082-024.A) on May 29, 2024

Inservice Testing (IST) (IP Section 03.01) (1 Sample)

(1) 1B-B RHR pump comprehensive performance test (1-SI-SXP-074-202.0) on April 10, 2024

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

(1) Unit 1 control air penetration, X-26B, local leak rate test (0-SI-SLT-032-258.1) on April 9, 2024

Ice Condenser Testing (IP Section 03.01) (1 Sample)

(1) Unit 1 ice condenser as left ice weighing (0-SI-MIN-061-105.0) on April 25, 2024

71114.06 - Drill Evaluation

Additional Drill and/or Training Evolution (1 Sample)

The inspectors evaluated:

(1) Emergency preparedness training drill including a general emergency resulting from a loss of coolant accident with fuel failure and potential containment failure conducted on May 15,

71124.01 - Radiological Hazard Assessment and Exposure Controls

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

(1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.

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

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

Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)

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

(1) Controls for items stored in the spent fuel pool

(2) Surveys of potentially contaminated material leaving the radiologically controlled area (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:

(1) Refueling water purification filter change out

(2) Core barrel movement

(3) Steam generator eddy current testing

(4) Fuel movement

High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (3 Samples)

The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):

(1) Tritiated Drain Collector Tank

(2) Liquid Radwaste Processing Area

(3) Unit 1 Lower Containment

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

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

71124.03 - In-Plant Airborne Radioactivity Control and Mitigation

Permanent Ventilation Systems (IP Section 03.01) (1 Sample)

The inspectors evaluated the configuration of the following permanently installed ventilation systems:

(1) Main control room emergency ventilation system

Temporary Ventilation Systems (IP Section 03.02) (1 Sample)

The inspectors evaluated the configuration of the following temporary ventilation systems:

(1) High Efficiency Particulate Air unit 700-8, seal table area

Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)

(1) The inspectors evaluated the licensees use of respiratory protection devices.

Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)

(1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.

71124.04 - Occupational Dose Assessment

Source Term Characterization (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.

External Dosimetry (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.

Internal Dosimetry (IP Section 03.03) (3 Samples)

The inspectors evaluated the following internal dose assessments:

(1) Positive whole body count on August 3, 2023

(2) In vitro analyses for tritium for diving operations during Spring 2023 outage

(3) Positive whole body count on April 17, 2024

Special Dosimetric Situations (IP Section 03.04) (2 Samples)

The inspectors evaluated the following special dosimetric situations:

(1) Declared pregnant worker on March 27, 2024

(2) Multiple dosimeter packs used for refueling canal work in 2023

71124.05 - Radiation Monitoring Instrumentation

Walkdowns and Observations (IP Section 03.01) (10 Samples)

The inspectors evaluated the following radiation detection instrumentation during plant walkdowns:

(1) Personnel contamination monitors, portal monitors, and small article monitors located at the radiologically controlled area RCA exit.

(2) Portable friskers located throughout the auxiliary building.

(3) Continuous air monitors located throughout the auxiliary building.

(4) Personnel contamination monitors located for use upon exiting Unit 1 containment.

(5) Portable alpha monitoring equipment staged at the spent fuel pool floor and containment exit.

(6) Area radiation monitors located throughout the auxiliary building.

(7) High purity Germanium and liquid scintillation detectors at the chemistry laboratory.

(8) Portable survey instruments such as telepoles and ion chambers staged for use throughout the plant and in the radiation protection (RP) office.

(9) Unit 1 lower containment high range area monitors, 1-RE-90-273 and 1-RE-90-274.

(10) Whole-body counters at dosimetry and in the RP office near the RCA exit.

Calibration and Testing Program (IP Section 03.02) (13 Samples)

The inspectors evaluated the calibration and testing of the following radiation detection instruments:

(1) Unit 1 upper containment post-accident high range area monitor 1-RE-90-271 calibrated September 13, 2022, and July 22, 2021

(2) Unit 1 lower containment post-accident high range area monitor 1-RE-90-274 calibrated April 19, 2021, and November 5, 2022

(3) CRONOS small article monitor 860780 calibrated February 26, 2024 and March 1, 2023

(4) Thermo Radeye PX neutron survey meter 952316 calibrated January 30, 2023, and February 27, 2024

(5) Thermo Radeye SX with Ludlum 43-92 alpha probe 952032 calibrated December 15, 2022, and February 1, 2024

(6) Ludlum 9-3 survey meter 860916 calibrated August 11, 2022, and June 8, 2023

(7) Mirion Telepole II TVA #951801 calibrated April 20, 2021, and March 18, 2024

(8) Ludlum 177 frisker TVA #951269 calibrated July 11, 2022, and August 22, 2023

(9) Ludlum 3030P alpha/beta counter TVA #951474 calibrated March 5, 2022, and March 21, 2024

(10) AMS-4 portable continuous air monitor TVA #860271 calibrated March 1, 2023, and April 21, 2024

(11) GEM-5 portal monitor TVA #14PM calibrated May 4, 2023, and March 8, 2024

(12) ARGOS personnel contamination monitors TVA #951453 calibrated August 23, 2022 and August 19, 2023

(13) Fuel pool area radiation monitor 0-RM-90-103 calibrated September 23, 2021, and May 9, 2023

Effluent Monitoring Calibration and Testing Program Sample (IP Section 03.03) (3 Samples)

The inspectors evaluated the calibration and maintenance of the following radioactive effluent monitoring and measurement instrumentation:

(1) Unit 1 shield building vent normal range noble gas monitor 1-RE-90-400 calibrated May 25, 2022, and August 17, 2023

(2) Waste disposal system liquid effluent monitor 0-RE-90-122 calibrated November 4, 2020, and May 4, 2022

(3) Unit 1 shield building vent post-accident high range area monitor 1-RE-90-261 July 25, 2022, and January 20,

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===

(1) Unit 1 (April 1, 2023, through March 31, 2024)

(2) Unit 2 (April 1, 2023, through March 31, 2024)

IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)

(1) Unit 1 (April 1, 2023, through March 31, 2024)

(2) Unit 2 (April 1, 2023, through March 31, 2024)

IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)

(1) Unit 1 (April 1, 2023, through March 31, 2024)

(2) Unit 2 (April 1, 2023, through March 31, 2024)

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

(1) April 1, 2023, through March 4, 2024.

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) Investigate cause of two recent steam dump valve stem shears (CR 1782471)

(2) Contingency actions used during unit 1 RCS drain down on April 14, 2024 (CR

===1924343)

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 Unit 2 Reactor Coolant Pump standpipe alarms that might be indicative of a more significant safety issue on June 29, 2024

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

(1) The inspectors evaluated the inadvertent isolation of shutdown cooling during a drain down of the reactor cavity on April 14, 2024

(2) The inspectors evaluated the RCS leak due to the lifting of a relief valve in the residual heat removal system in Mode 5 on April 24,

INSPECTION RESULTS

Failure to Survey for Airborne Radioactivity Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.8] - 71124.01 Radiation Safety NCV 05000327,05000328/2024002-01 Procedure Open/Closed Adherence An NRC-identified Green NCV of TS 5.4.1, Procedures, was identified for the licensee's failure to implement an air-sampling procedure during radiological protection job coverage of work in the Unit 1 containment keyway. Specifically, the licensee failed to obtain an air sample while performing work in an area with removable contamination levels greater than 100,000 disintegrations per minute (dpm) per 100 cm2.

Description:

On April 17, 2024, during a refueling outage, several workers entered the Unit 1 containment keyway to demobilize a variety of equipment (e.g. scaffold breakdown and removal). Radiological surveys showed removable surface contamination levels in this area of up to 1,000,000 dpm of beta-gamma activity per 100 cm2. The workers reported dusty conditions during movement of equipment in the area, indicating that some of the removable contamination may have become resuspended in the air. After the work was complete, four of the workers had difficulty clearing the contamination monitors at the radiologically controlled area exit. After changing scrubs and/or showering, all four were administered whole-body counts (WBCs), which came back positive for internal contamination. Upon reviewing the job coverage survey from this event, the inspectors identified that the licensee failed to take an air sample in the area while the work was being performed. A follow-up air sample obtained approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the event showed an airborne radioactivity concentration of 0.15 derived air concentration (DAC). Although this indicated that airborne particulate radioactivity was still present in the area, it was likely not an accurate assessment of the true conditions during the event, thereby degrading the licensee's ability to evaluate the workers' intakes. However, multiple follow-up WBCs showed that the intakes were not significant, and the licensee was able to bound the dose estimate in each case to less than 10 mrem committed effective dose equivalent (CEDE). In addition, there was no significant alpha contamination present which could have invalidated the WBC results. The inspectors noted that licensee procedure NISP-RP-003, Radiological Air Sampling, requires an air sample to be taken during work in areas with removable beta-gamma contamination levels greater than 100,000 dpm per 100 cm2.

Corrective Actions: A follow-up air sample was taken of the work area after the positive whole-body counts occurred. The workers were instructed to report for follow-up whole-body counts to adequately assess the magnitude of the uptakes.

Corrective Action References: CR 1925158

Performance Assessment:

Performance Deficiency: The inspectors determined that the failure to obtain an air sample while performing work in an area with smearable contamination levels greater than 100,000 dpm per 100 cm2, as required by a licensee procedure, 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 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 failure to obtain accurate air samples in areas with significant removable surface contamination levels can potentially impact the licensee's ability to assess dose to workers should an intake event occur.

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: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. A cross-cutting aspect of [H.8], Procedure Adherence, was assigned because the performance deficiency stemmed directly from a failure to follow a procedure.

Enforcement:

Violation: Technical Specification 5.4.1 requires procedures recommended by Regulatory Guide 1.33, Revision 2, be established, implemented, and maintained, including procedures for airborne radioactivity monitoring. Licensee procedure NISP-RP-003, Radiological Air Sampling, states, in part, that work in areas with removable beta-gamma contamination levels greater than 100,000 DPM per 100 cm2 requires air sampling. Contrary to this, on April 17, 2024, work was performed in an area with removable beta-gamma contamination levels greater than 100,000 DPM per 100 cm2 and no air sample was taken.

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

Inadvertent isolation of shutdown cooling Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.7] - 71153 NCV 05000327/2024002-02 Documentation Open/Closed A self-revealed Green finding and associated non-cited violation of Technical Specification (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement written procedures for activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Specifically, the licensee failed to establish and maintain adequate procedures for reactor coolant system (RCS) drain and fill operations. As a result, main control room operators unintentionally isolated shutdown cooling (SDC) flow when responding to a loss of communication with a remote valve operator during reactor cavity drain down.

Description:

Single train Shutdown Cooling (SDC) alignment of Residual Heat Removal (RHR) is restricted to discharge into only two RCS cold legs. Normally, either the A RHR train uses the A RHR pump injecting into the RCS loops 2 and 3 cold legs or the B RHR train uses the B RHR pump injecting into the RCS loops 1 and 4 cold legs. The discharge headers can be cross connected through the RHR crosstie valves (1-FCV-74-35 and 1-FCV-74-33). One or both crosstie valves are normally closed for single train SDC operation. If desired, both crosstie valves may be opened to use the A RHR pump to inject into RCS loops 1 and 4 cold legs or B RHR pump to inject into the RCS loops 2 and 3 cold legs. The RHR system refueling water return valve, 1-HCV-74-34, taps off the RHR system between the RHR crosstie valves. If 1-HCV-74-34 is open, closing 1-FCV-74-35 or 1-FCV-74-33 isolates the RHR flow path to the refueling water storage tank (RWST) from A or B train, respectively. When the SDC trains are in a normal single train lineup (no SDC flow thru the cross ties), shutting RHR crosstie valves can isolate the flow path to the RWST without affecting cooling flow to the RCS.

On April 14, 2024, operations personnel were in the process of draining the reactor cavity to the RWST in preparation for reactor head set during refueling outage U1R26. SDC was in operation with the 1B residual heat removal (RHR) pump injecting through the RHR crosstie valves (1-FCV-74-35 and 1-FCV-74-33), and into cold legs 2 and 3. Drain down to the RWST was controlled by a remote valve operator who was locally operating 1-HCV-74-34. In this cross tied configuration, closing 1-FCV-74-35 or 1-FCV-74-33 (RHR crosstie valves) isolates SDC flow to the RCS. Sequoyah procedure 0-GO-13, Reactor Coolant System Drain and Fill Operation, did not note the risk of using the RHR cross tie valves during a drain down and did not specify that this lineup was not preferred, only that the RHR system should be in a ONE TRAIN cooldown mode of operation.

At 02:38, with a cavity level approximately 21 feet above the reactor flange, the main control room (MCR) lost communications with the remote valve operator at 1-HCV-74-34. In accordance with 0-GO-13, section 5.4.2 for lost communications, MCR operators closed 1-FCV-74-35 and 1-FCV-74-33. Closing 1-FCV-74-35 and 1-FCV-74-33 isolated the RHR flow path to the RWST, stopping the drain down as intended, but because SDC was in a cross tied configuration, this also isolated SDC flow.

Unit 1 entered Tech Spec LCO 3.9.6 condition B since RHR flow was reduced to <2000 gpm with water level <23 ft above the top of the reactor vessel flange. At 02:42, MCR communication with the remote valve operator was restored and the operators in the MCR reopened the crosstie valves and restored SDC flow to >2000 gpm. Unit 1 exited LCO 3.9.6 condition B.

Corrective Actions: Sequoyah Unit 1 restored SDC flow as soon as communications between the MCR and the 1-HCV-74-34 operators was restored, and it was confirmed that 1-HCV-74-34 was shut.

RHR alignment was changed for the remainder of the drain down operation such that the cooling flow to the RCS was not required to go through the RHR cross connect valves. In this configuration drain down could be stopped from the MCR without reducing SDC flow to the RCS.

Sequoyah has corrective actions to make procedure changes which require redundant communications between MCR and the refueling water return valve during drain down operations and will also add notes to the procedure that describe the higher risk of using the RHR crosstie during RCS drain down.

Corrective Action References: CR 1924343

Performance Assessment:

Performance Deficiency: The inspectors determined that the licensees failure to adequately establish and implement operating procedures for draining the reactor cavity that were appropriate to the circumstance was a performance deficiency reasonably within the licensees ability to foresee and prevent. Specifically, the licensee procedure 0-GO-13, Reactor Coolant System Drain and Fill Operation, governing operation of the RHR system during drain down operations did not note the risk of using the RHR cross tie valves during a drain down. The inadequate procedure resulted in operators unintentionally isolating SDC when they secured drain down flow in a manner that was allowed by the procedure.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to adequately establish and maintain adequate procedures for RCS drain and fill operations resulted in the unplanned loss of decay heat removal.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Safety SDP. The inspectors reviewed IMC 0609 Appendix G, 1, Shutdown Operations Significance Determination Process, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding affects the Initiating Events cornerstone. Exhibit 2, Initiating Events Screening Questions, directs the use of Appendix G Phase 2 screening since the loss of RHR did not occur when the refuel canal/cavity was flooded.

A regional Senior Reactor Analyst conducted a Phase 2 evaluation in accordance with IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, dated January 8, 2020. The finding was determined to represent a late outage time window Loss of Residual Heat Removal (LORHR) precursor that occurred in Plant Operational State (POS) 2. The analysis did not credit the volume of water in the refueling cavity above the upper internals and did not include any additional operator recovery actions that were not explicitly addressed in the screening guidance. Worksheet 9 of 2 was used to assess the significance of the finding. The conditional core damage probability sequences of interest included an unrecovered loss of Residual Heat Removal (RHR) accompanied by failure of makeup to the refueling water storage tank, and loss of RHR with failure to feed the reactor pressure vessel. The Phase 2 analysis yielded conditional core damage probability sequences that were less than E-06 which screened the finding to green significance. The significance of the finding was mitigated by the availability of plant equipment that would be necessary to mitigate a loss of RHR flow and late time window lower decay heat which afforded more time for implementation of mitigative operator actions.

Cross-Cutting Aspect: H.7 - Documentation: The organization creates and maintains complete, accurate and up-to-date documentation. Specifically, the licensee procedure 0-GO-13, Reactor Coolant System Drain and Fill Operation, did not note the risk of using the RHR cross tie valves during a drain down and did not specify that this lineup was not preferred, only that the RHR system should be in a ONE TRAIN cooldown mode of operation. The inadequate procedure resulted in operators unintentionally isolating SDC when they secured drain down flow in a manner that was allowed by the procedure.

Enforcement:

Violation: Sequoyah Unit 1 TS 5.4.1.a requires that written procedures shall be established, implemented, and maintained as recommended by Appendix A of Regulatory Guide 1.33, Revision 2. Appendix A section 3.a and 3.c requires procedures for Startup, Operation, and Shutdown of the Reactor Coolant System and Shutdown Cooling System, respectively.

Contrary to the above, on April 14, 2024, the licensee failed to establish appropriate procedures for safe operation of the Shutdown Cooling System during RCS drain down operations. Sequoyah procedures allowed for an RHR system lineup that results in losing Shutdown Cooling flow to the RCS if drain down flow is secured by the Main Control Room during a communication loss event with remote valve operators. This condition led to Sequoyah Unit 1 losing Shutdown Cooling flow to the RCS for approximately 3 minutes.

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 July 22, 2024, the inspectors presented the integrated inspection results to Tom Marshall and other members of the licensee staff.

• On April 11, 2024, the inspectors presented the ISI Exit Meeting inspection results to Tom Marshall - Site VP and other members of the licensee staff.

• On April 26, 2024, the inspectors presented the Radiation Protection Inspection Exit Meeting inspection results to T. Marshall and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Corrective Action CR 1922336 Enhance AOP-N.02 Tornado Watch/Warning 04/05/2024

Documents

Procedures AOP-N.02 Tornado Watch/Warning Revision 46

71111.04 Procedures 0-GO-13 Reactor Coolant System Drain and Fill Operations Revision 111

0-SO-74-1 Residual Heat Removal System Revision 116

0-SO-82-3 Standby Mode Parameters Revision 75

0-SO-82-3 Verification of Available Standby Condition Revision 75

Seciton 5.2

71111.05 Fire Plans CON-0-685-00 Pre-Fire Plan - Control Building (Elevation 685) Revision 8

DGB-0-722-00 Pre-Fire Plan - Diesel Generator Building (Elevation 722) Revision 7

RXB-0-679-01 Pre-Fire Plan - Reactor Building (Elevation 679) Revision 3

RXB-0-701-01 Pre-Fire Plan - Reactor Building Annulus Area (Elevations Revision 3

701 & 721)

RXB-0-734-01 Pre-Fire Plan - Reactor Building (Elevation 734) & Annulus Revision 3

(Elevations 740, 759, & 778)

71111.11Q Procedures 0-GO-13 Reactor Coolant System Drain and Fill Operations Revision 111

71111.12 Corrective Action CR 1926802 Prompt Investigation for 1A RHR discharge valve relief valve 04/24/2024

Documents lifting

Work Orders WO 124452055 Disassemble, inspect, refurbish and test Crosby relief valve 04/29/2024

RV-1-8856A

71111.13 Miscellaneous High Risk Unit 1 Down Power to 75% to support 500kV switching and 06/11/2024

Management 1C CBP Degraded Oil Pressure repair

Plan - WO 24550851

Procedures 0-GO-16 System Operability Checklist Revision 37

NPG-SPP-07.3.4 Protected Equipment Revision 15

Work Orders WO 124550851 Breaker Disagreement on 5074 (Bradley Line) 06/13/2024

71111.15 Corrective Action CR 1922292 1A Containment Spray Heat Exchanger obstructed tubes 04/05/2024

Documents CR 1929467 Abnormally high oil level in the U2 TDAFW Pump lubricating 05/06/2024

oil sight glass

Corrective Action CR 1923113 NRC identified areas of peeling coatings in lower 04/09/2024

Inspection Type Designation Description or Title Revision or

Procedure Date

Documents containment

Resulting from

Inspection

Engineering Engineering PCR 1925285 Urgent change to 1-SI-OPS-068-001.0 (Low 04/18/2024

Evaluations Evaluation for Temperature Overpressure Protection) to allow 3 inch vent

WO 124439095 path utilizing one PORV and Block Valve open and the

pressurizer vent valve unflanged and tagged open to comply

with TS 3.4.12.b.

Joseph Oat Heat Exchanger Damage Evaluation for J-7162 Containment Revision 0

Corporation Spray HX 1A

Report No: TM-

291

Structural 1RF26 Risk Assessment of Degraded 1A Containment Revision 0

Integrity Spray HX

Associates, Inc

Report

2300527.401

Miscellaneous Doc. No. 603461- Structural Technologies TVA Sequoyah Nuclear Plant Unit 1 Revision 1

RPT-001 Lower Containment Slab Condition Assessment

Operability 2048-0045-RPT- Turbine Driven Auxiliary Feedwater Pump Turbine Oil Water Revision 0

Evaluations 001 Intrusion Event at Watts Bar Nuclear Plant Unit 1. Past

Operability Evaluation

PDO for CR Prompt Determination of Operability Documentation for CR 05/08/2024

29467 1929467

71111.18 Miscellaneous SO-24-009 ACMP 2A Safety Injection Pump Oil Particulates 02/22/2024

Work Orders WO 124365316 Fabricate and Install TMOD SQN-2-2024-063-01 05/21/2024

71111.20 Corrective Action CR 1924343 0-GO-13 contingency actions used during RCS draindown 04/14/2024

Documents on 4/14/24

Corrective Action CR 1922249 NRC identified white deposits on Aux building wall in the 04/05/2024

Documents vicinity of the spent fuel pool

Resulting from CR 1922352 NRC identified residue in U1 raceway ceiling 04/05/2024

Inspection CR 1923113 NRC identified loose/peeling coating in lower containment 04/09/2024

CR 1926775 Boron buildup on 1-VLV-68-580 04/24/2024

CR 1927174 Legacy scaffold members in Unit 1 containment 04/26/2024

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.24 Corrective Action CR 1924493 1A RHR pump diff pressure outside acceptance criteria due 04/15/2024

Documents to new pump installation

CR 1924698 Late CR on 1A RHR failed acceptance criteria 04/16/2024

Procedures 0-PI-SLT-063- Leak Check of ECCS Relief Valve Header Revision 11

200.0

0-SI-SXI-000- ASME Section XI Inservice Pressure Test Revision 27

201.0

1-SI-SXI-068- Leakage Test of the Reactor Coolant Pressure Boundary Revision 18

201.0

Work Orders WO 123385617 U1 Ice Condenser - Ice Weighing (as left) 04/27/2024

71124.01 Corrective Action CR 1925158 04/17/2024

Documents

Engineering Sequoyah Nuclear Plant Alpha Characterization Report 03/20/2024

Evaluations

Procedures NISP-RP-003 Radiological Air Sampling Rev. 1

NPG-SPP-05.1.1 Alpha Radiation Monitoring Program Revision 10

Radiation 041024005 669' WCT/RWPP Areas - Air Sample 04/10/2024

Surveys SQN-M- U1 Equipment Pit Gripper Inspection 03/21/2024

240321-6

SQN-M- Keyway Equipment Removal Survey 04/17/2024

240417-10

71124.03 Procedures 0-PI-FPU-049- Self Contained Breathing Apparatus Revision 33

401.M

71152A Corrective Action CR 1924343 0-GO-13 contingency actions used during RCS drain down 04/14/2024

Documents on 4/14/24

CR 1925215 4.0 Critique for Operations drain to midloop 04/18/2024

CR 1937403 Potential System Impacts to Previous Steam Dump Failures 06/12/2024

Engineering BE 18954 Central Labs Analysis of Steam Dump SQN-2-FCV-001- 10/20/2023

Evaluations 0104, -0111, -0112

71152S Corrective Action CR 1936881 Received 2-XA-55-5B window (B-2) "LS-62-19A REAC 06/11/2024

Documents COOL PMP 2 STANDPIPE LVL HIGH-LOW"

CR 1939614 Received unexpected annunciator on 2-M-5B Window B2 06/24/2024

"LS-62-19A REAC COOL PMP 2 STANDPIPE LVL HIGH

LOW"

Inspection Type Designation Description or Title Revision or

Procedure Date

CR 1941069 Low level alarm in RCP #2 standpipe 06/29/2024

Work Orders WO 122976924 MEG/CM2-LS-62-19B/ Inspect and adjust/repair/replace 10/10/2024

level switches on U2 #2 RCP standpipe as required.

71153 Corrective Action CR 1924343 0-GO-13 contingency actions used during RCS draindown 04/14/2024

Documents on 4/14/24

CR 1926579 1A RHR discharge relief valve downstream of 63-93 lifted 04/24/2024

CR 1926614 Pressurizer thermal limits exceeded during leak response 04/24/2024

CR 1926698 PCR for AOP-R.03 on PRT leakage and double disk leakoff 04/24/2024

CR 1927286 PCR for AOP-R.05 - Add steps to address PRT conditions to 04/27/2024

prevent rupture

Corrective Action CR 1926787 Engineering analysis due to PZR spray line temperature 04/24/2024

Documents differential exceeding 320 degrees

Resulting from

Inspection

20