IR 05000483/2023004

– Integrated Inspection Report 05000483/2023004

ML24017A294
Person / Time
Site:	Callaway
Issue date:	01/19/2024
From:	Greg Werner NRC/RGN-IV/DORS/PBB
To:	Diya F Ameren Missouri
References
IR 2023004
Download: ML24017A294 (36)

Text

January 19, 2024

SUBJECT:

CALLAWAY PLANT - INTEGRATED INSPECTION REPORT 05000483/2023004

Dear Fadi Diya:

On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Callaway Plant. On January 9, 2024, the NRC inspectors discussed the results of this inspection with J. Beck, Vice President Nuclear Engineering, and other members of your staff. The results of this inspection are documented in the enclosed report.

Four findings of very low safety significance (Green) are documented in this report. All 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.

A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as a non-cited violation 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at the Callaway Plant.

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 IV; and the NRC Resident Inspector at the Callaway Plant.

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, Gregory E. Werner, Chief Reactor Projects Branch B Division of Operating Reactor Safety Docket No. 05000483 License No. NPF-30

Enclosure:

As stated

Inspection Report

Docket Number: 05000483

License Number: NPF-30

Report Number: 05000483/2023004

Licensee: Ameren Missouri

Facility: Callaway Plant

Location: Steedman, MO

Inspection Dates: October 1 through December 31, 2023

Inspectors: C. Alldredge, Health Physicist B. Baca, Senior Health Physicist N. Brown, Resident Inspector J. Drake, Senior Reactor Inspector J. Mejia, Reactor Inspector J. O'Donnell, Senior Health Physicist S. Schwind, Senior Resident Inspector

Approved By: Gregory E. Werner, Chief Reactor Projects Branch B Division of Operating Reactor Safety

Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Callaway Plant, 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 71124.0

List of Findings and Violations

Failure to Meet Surface Preparation Requirements for Volumetric Examinations in Accordance with ASME Section XI Code Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71111.08P NCV 05000483/2023004-01 Open/Closed The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of 10 CFR 50.55a, Codes and Standards, involving the licensees failure to meet examination surface requirements for volumetric examinations in accordance with ASME Section XI Code.

Failure to Follow Maintenance Procedure Regarding Post-Maintenance Testing of the Control Room Envelope Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.8] - 71111.24 NCV 05000483/2023004-02 Procedure Open/Closed Adherence The inspectors identified a Green finding and associated non-cited violation of Technical Specification 5.4.1a. when the licensee failed to follow the applicable post-maintenance test procedure after replacing the seal on the control room door.

Failure to Identify and Correct an Inoperable Containment Instrument Sump Level Indicator Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.5] - 71153 Systems NCV 05000483/2023004-04 Operating Open/Closed Experience The inspectors reviewed a self-revealed Green finding and associated non-cited violation of 10 CFR Part 50 Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, from March 2 through September 30, 2023, the licensee failed to identify and correct an inoperable containment instrument sump level detector. As a result, the containment sump level and flow monitoring system was inoperable for greater than the technical specification allowed outage time. This event was reported as License Event Report 05000483/2023-001-00.

Inadequate Procedure Results in a Fire in the Secondary Alarm Station Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.8] - 71153 NCV 05000483/2023004-03 Procedure Open/Closed Adherence The inspectors reviewed a self-revealed Green, non-cited violation of License Condition 2.C.(5), Fire Protection Program, regarding an inadequate fire prevention program. Specifically, the licensees fire prevention program did not address overloading electrical outlets which resulted in a fire in an outlet in the secondary alarm station.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000483/2022-003-01 LER 2022-003-01 for 71153 Closed Callaway Plant Unit 1,

Class 1E Electrical Air Conditioning System Thermal Expansion Valve Failure Resulted in Condition Prohibited by Technical Specifications LER 05000483/2023-001-00 LER 2023-001-00 for 71153 Closed Callaway, Unit 1, Inoperable Instrument Tunnel Sump Level Indication Resulted in Condition Prohibited by Technical Specifications

PLANT STATUS

The Callaway plant began the inspection period at rated thermal power. On September 29, 2023, the licensee commenced a planned shutdown in preparation for Refueling Outage 26.

The plant was taken offline and remained in a shutdown condition from September 30 through November 5, 2023, when plant startup was commenced. The main generator was synchronized to the grid on November 7, 2023, and full power operations resumed on November 9, 2023. The plant remained at rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

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

71111.01 - Adverse Weather Protection

Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures on November 1, 2023.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (2 Samples)

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

(1) residual heat removal train A on October 5, 2023

(2) turbine-driven auxiliary feedwater steam supply following maintenance on FCST0001, turbine-driven auxiliary feedwater pump steam supply drain trap, on November 14, 2023

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 Samples)

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

(1) fire area C-9, engineered safety features switchgear on October 11, 2023

(2) fire area D-1, diesel generator A on October 11, 2023

(3) fire area RB-1, reactor building on October 23, 2023

(4) fire area A-13, auxiliary feedwater pump room B, fire area A-14, auxiliary feedwater pump room A, and fire area A-15, turbine-driven auxiliary feedwater pump room on November 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 9 through October 20, 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 Examination

• safety injection, 2-EM-03S014-C, pipe to elbow

• main steam, 2-AB-01-F050, pipe to pipe

• CVCS, 2-BG-22-F001, nozzle to elbow

• RPV Head Closure Bolting, 1-54, studs

Magnetic Particle Examination

• essential service water, EF-009-HBC-30, weld overlay repair

Dye Penetrant Examination

• pipe support, EJ02C022131, lug welds

Visual Examination

• fuel pool CLG HX B, EEC01B, inner shell side

• pipe support, EJ01-R017/134, support welds

• RPV Head Closure Bolting, 1-54, nuts and washers

Welding Activities

• gas tungsten arc welding and shielded metal arc welding o feedwater, AE-079-EBB-l4, FW-02, FW-03, elbow to pipe

PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection

Activities (IP Section 03.02) (1 Sample)

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

(1)

• The inspectors observed portions of the reactor vessel head inspection, penetration nozzles 1, 2, 5, 6, 8, 32, 33, 36-40, 42, 44, 45, 47, 50-53, 56-58, 61, 66, 68, 71, 72, 75, RVLIS C-B, and general head area.

PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)

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

(1)

• 202207467, pass hot leg 1 sample out containment isolation valves downstream test valve, SJV0056

• 202207698, threaded connection, Cat. 1, Class 1, FLM 17-0002, BGFT0145

• 202207879, packing, Cat. 1, Class 2, FLM 10-0011, BGV0102

• 202303556, reactor coolant drain tank heat exchanger out header lo side flush conn pipe cap, HBC0033

• 202307476, train A walkdown, October 15, 2023 PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04)

• No steam generator tube inspections were scheduled nor performed during this outage.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

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

(1) The inspectors observed and evaluated licensed operator performance in the control room during reactor startup on November 5, 2023.

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

(1) The inspectors observed and evaluated reactor startup and steam generator faults training in the simulator on November 16, 2023.

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 remain capable of performing their intended function:

(1) September 28, 2023, failure of SGK05A, train A class 1E air conditioner on November 9, 2023

71111.13 - Maintenance Risk Assessments and Emergent Work Control

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

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

(1) risk profile and risk management actions for Refueling Outage 26 on October 1, 2023

(2) risk management action for train A spent fuel pool decay heat removal yellow risk condition on October 16, 2023

(3) risk management actions for drain down to lowered inventory for reactor head installation on October 19, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (2 Samples)

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

(1) instrument tunnel sump level indicator on October 1, 2023

(2) essential service water line EF-007-HBC below minimum required wall thickness on October 10, 2023

71111.20 - Refueling and Other Outage Activities

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

(1) The inspectors evaluated various Refueling Outage 26 activities from September 29 through November 5, 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) (14 Samples)

(1) SGK05A, class 1E electric equipment air conditioning unit on October 2, 2023

(2) replacement of ESFAS channel 4 logic card SA036E on October 7, 2023

(3) install in-line power supply filter in nuclear instrument rack SE054D on October 10, 2023

(4) containment recirculation sump to residual heat removal pump B isolation valve EJ-HV-8811B post maintenance test on October 10, 2023

(5) residual heat removal train A accumulator injection supply header relief valve (EJ8856A) replacement on October 11, 2023

(6) emergency diesel generator B following fuel oil system inspections and maintenance on October 12, 2023

(7) VT-2 inspection following replacement of valve EFHV0038, essential service water train B to ultimate heat sink cooling tower on October 13, 2023

(8) control room door seal replacement on October 15, 2023

(9) containment cooler C preventative maintenance test on October 15, 2023

(10) essential service water train A to containment air coolers outlet valve EF-HV-0031 post-maintenance test on October 17, 2023

(11) safety injection accumulator C check valve BB8948C repairs on November 5, 2023

(12) body-to-bonnet leak repairs on BLV0007, reactor makeup water storage tank recycle/waste evaporator condensate inlet isolation valve BLV0007 on November 9, 2023

(13) overhaul/rebuild steam trap FCST0001, turbine-driven auxiliary feedwater pump steam supply drain trap on November 13, 2023

(14) centrifugal charging pump A motor oil sample and change out on November 16, 2023

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) BGHV8160 inservice test on October 17, 2023

(2) reactor coolant system resistance temperature detector cross-calibration on October 25, 2023

(3) steam generator A narrow range level channel 4 operability test on November 27, 2023

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

(1) EGHV0059, component cooling water from reactor coolant system on October 5, 2023

71114.06 - Drill Evaluation

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

The inspectors evaluated:

(1) emergency preparedness drill on November 16,

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) surveys of potentially contaminated material leaving the radiologically controlled area

(2) workers exiting the containment building during a refueling outage

Radiological Hazards Control and Work Coverage (IP Section 03.04) (3 Samples)

The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:

(1) eddy current testing of flux thimbles using radiation work permit R26STEC

(2) upper internals move using radiation work permit R26UPPERINLIFT

(3) access to regenerative heat exchanger cocoon for snubber testing using radiation work permit 22505893500

High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)

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

(1) (VHRA) instrument in-core tunnel (2) (HRA) 7122 - waste holdup tank room (3) (HRA) 7123 - primary waste evaporator bottoms tank room (4) (HRA) 7213 - entrance to stock system/decant tank room (5) (HRA) 7220 - valve room (primary resin storage tank)

Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (2 Samples)

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

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

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) (1 Sample)

The inspectors evaluated the following internal dose assessments:

(1) April 23, 2022, incident where an individual accidentally fell in the reactor cavity pool

Special Dosimetric Situations (IP Section 03.04) (1 Sample)

The inspectors evaluated the following special dosimetric situations:

(1) Form 5s for five declared pregnant workers

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) October 2022 through October 2023

IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (1 Sample)

(1) October 2022 through October 2023

IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (1 Sample)

(1) October 2022 through October 2023

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

(1) January 1, 2022, through June 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) January 1, 2022, through June 30, 2023

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

(1) July 1 through December 31, 2023. The inspectors reviewed the licensees corrective action program for potential adverse trends that might be indicative of a more significant safety issue. The inspectors observed a negative trend in worker protection assurance during maintenance activities.

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

(1) The inspectors evaluated a fire in the secondary alarm station on October 1, 2023.

(2) The inspectors evaluated safety injection accumulator C check valve BB8948C that failed inservice leak rate testing on October 27, 2023.

Event Report (IP Section 03.02) (2 Samples)

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

(1) The inspectors reviewed Licensee Event Report 2023-001-00, for Callaway Plant, Unit 1, Inoperable Instrument Tunnel Sump Level Indication Resulted in Condition Prohibited by Technical Specifications (ADAMS Accession No. ML23333A141). The circumstances surrounding this LER are documented in the Inspection Results section of this report. This LER is Closed.

(2) The inspectors reviewed Licensee Event Report 050000483/2022-003-01 for Callaway, Unit 1, Class 1E Electrical Air Conditioning System Thermal Expansion Valve Failure Resulted in Condition Prohibited by Technical Specifications (ADAMS Accession No. ML23194A172). The inspection conclusions associated with this LER are documented in Inspection Report 2023001 (ML23102A139). This LER is Closed.

INSPECTION RESULTS

Failure to Meet Surface Preparation Requirements for Volumetric Examinations in Accordance with ASME Section XI Code Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71111.08P NCV 05000483/2023004-01 Open/Closed The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation (NCV) of 10 CFR 50.55a, Codes and Standards, involving the licensees failure to meet examination surface requirements for volumetric examinations in accordance with ASME Section XI Code.

Description:

During refueling outage 26 the licensee identified surface defects on the Class 1 pressurizer safety relief piping below the elbow to flange welds at the inlets to the safety relief valves. The licensee initiated condition report 202307267 attributing the cause to usage of a hydraulic torque tool that tensioned the safety relief valve fasteners where the reaction arm would contact the pipe surface. This contact created surface irregularities around the piping surface and would have happened during pressurizer safety relief valve replacement which occurred between 1984 and 2008 as described in the condition report. After 2008, the hydraulic torque tool was updated to a newer version which allowed the reaction arm to contact the adjacent fastener, eliminating contact with piping surfaces.

After discovery of the indications, the licensee photographed, measured the surface indications depths, and contracted with a vendor to perform an analytical evaluation of the surface indications. The evaluation determined that the indications did not require immediate corrective actions as the minimum wall thickness has adequate margin and was acceptable for 18 months of continued service (until the next refueling cycle). The evaluation also recommended the removal of the indications stress risers and performance of non-destructive examination after repair. The licensee created work orders to perform repairs and non-destructive examination for refueling outage 27.

The licensee performs risk informed volumetric examinations once every interval on the elbow to flange welds of the inlets to the pressurizer safety relief valves. The volumetric examination procedure specifies the required examination volume shall be, as a minimum, the lower 1/3 weld volume and base material for a distance of 0.25 from each weld toe. The procedure also specifies limitations or other conditions, i.e., surface condition, that prevent a complete examination of the required volume shall be documented and the owner should be made aware of scanning limitations at the earliest possible opportunity. The procedure provides surface condition requirements where the examination surface shall be free of irregularities which may interfere with ultrasonic wave transmission.

The inspectors identified, through review of photos of each of the three elbow to flange welds and by visual inspection, that the surface conditions of each line also included portions of the surface irregularities within the required examination surface and volume. The inspectors further reviewed an examination report, dated May 6, 2010, from the 3rd interval which included a sketch of the examination volume and risk informed volume that extended the examination volume 0.5 distance from the weld toe. The ultrasonic examination reports dated May 6, 2010, and May 23,2019, for both the 3rd and 4th intervals did not document the surface irregularities of the pipe condition. The weld surface condition was documented as smooth in the 2010 exam report and ground flush in the 2019 exam report.Section XI, of ASME code, Subsection IWB, Figure IWB-2500-8, defines the examination volume to be 0.25 from the weld end (toe) and the examination surface to be 0.5 from the weld end (toe).

Many of the surface irregularities were raised above the pipe surface outer diameter. Surface irregularities cause interference with ultrasonic wave transmission and the transducers scan path.

Corrective Actions: The licensee documented the issue in their corrective action program and initiated actions to remediate the indications in the next outage.

Corrective Action References: condition report 202308929 was initiated to further evaluate the volumetric examinations previously performed.

Performance Assessment:

Performance Deficiency: The licensees failure to meet surface preparation requirements in accordance with ASME Section XI was a performance deficiency. Specifically, the licensee failed to meet surface preparation requirements during the 2010 and 2019 ultrasonic examinations.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the RCS Equipment and Barrier Performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure of the surface to be free of irregularities could lead to not detecting flaws. The inspectors screened the finding using IMC 0609, Attachment 4, Initial Characterization of Findings with Table 2. The Barrier Integrity Cornerstone Note directs all other RCS boundary issues will be considered under the Initiating Events Cornerstone. Inspection Manual Chapter 0609, Attachment 4, Table 3, routes the finding to IMC 0609, Appendix A, because the inspectors answered all the SDP Appendix Router questions no.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding screened as very low safety significance (Green) because the inspectors answered all IMC 0609, Appendix A, Exhibit 1, Initiating Events Questions no.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: Title 10 CFR Part 50.55a(g)(4), Inservice Inspection Standards Requirement for Operating Plants, requires that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirement set forth in Section XI of editions and addenda of the ASME Boiler and Pressure Vessel Code. ASME Section XI, Article III-4000 Examination, Subsection III-4200, Surface Preparation, states, in part, that the examination surface shall be free of irregularities which interfere with ultrasonic wave transmission.

Figure IWB-2500-8 describes the examination surface as 0.5 from the weld end and the examination volume as 0.25 from the weld end.

Contrary to the above, during the code required examinations in 2010 (3rd ASME Section XI Code Interval), and in 2019 (4th ASME Section XI Code Interval), the licensee failed to ensure the examination surface was free of surface irregularities which interfere with ultrasonic wave transmission.

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 Maintenance Procedure Regarding Post-Maintenance Testing of the Control Room Envelope Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.8] - 71111.24 NCV 05000483/2023004-02 Procedure Open/Closed Adherence The inspectors identified a Green finding and associated non-cited violation of Technical Specification 5.4.1a. when the licensee failed to follow the applicable post-maintenance test procedure after replacing the seal on the control room door.

Description:

On October 13, 2023, while in Mode 6 (Refueling) the licensee documented a failure of the seal on inner control room door DSK 36043 in condition report 202307412. This is one of two doors in series used for normal access to the control room. The other, door DSK 36042, is a hardened door which is normally left open. Either door can be used as a pressure boundary to maintain the integrity of the control room envelope and operability of the control room emergency ventilation system (CREVS). The CREVS is not required to be operable in Mode 6 unless irradiated fuel assemblies are being moved; nevertheless, the licensee declared door DSK 36043 inoperable and established administrative controls to close door DSK 36042, if necessary, to maintain operability of CREVS. Job 23003787 was initiated to replace the door seal which was completed on October 14 at 12:43 am.

Job 23003787 did not specify any post-maintenance test or quality control inspections.

On October 14 at 3:24 am the licensee began reloading the reactor core which required CREVS to be operable due to the movement of irradiated fuel assemblies. Control room door DSK 36042 was still under administrative control, so CREVS operability was not challenged. However, door DSK36043 was declared operable on October 14 at 11:21 am and those administrative controls were discontinued. At this point, door DSK 36043 would have been relied upon as a pressure boundary in the event of a fuel handling accident. Core alterations continued until completed October 16 at 3:00 am.

On October 16, following completion of core alterations, the licensee wrote condition report 202307517 to document that no post-maintenance test had been performed following corrective maintenance on door DSK 36043. This condition report stated, Door was inspected and meets all the requirements of OSP-KC-00015 [Fire Door Inspections] for a Fire Door. Either door DSK 36043 or door DSK 36042 can satisfy the pressure boundary function and may be open under administrative control. No other tests or inspections to verify the pressure function of the door were performed and no administrative controls were established to close door DSK 36042, if necessary. Operability of CREVS during core alterations was not questioned in the condition report.

On October 23, the inspectors discussed with the licensee the lack of a pressure boundary test or inspection following maintenance on door DSK 36043 and the lack of administrative controls on door DSK 36042. A subsequent pressure boundary inspection on door DSK 36043 indicated an unacceptable gap in the door seal of approximately 8.2 square inches which resulted in the door being declared inoperable. The seal was replaced again on October 24 under job 23004002. All required tests and inspections were completed satisfactorily, and the door was declared operable.

Procedure APA-ZZ-0000750, Hazard Barrier Program, designates door DSK 36043 as a pressure boundary. Procedure APA-ZZ-00322, Appendix E, Post-Maintenance Test Program, requires that following seal replacement, pressure boundary doors are to be inspected according to Procedure MPM-SK-QW001, Addendum 5, Service and Inspection of Pressure Doors. This inspection was not performed following the initial corrective maintenance on October 14.

Following replacement of the door seal on October 24, the licensee evaluated the 8.2 square inch gap and concluded that the control room envelope remained operable since CREVS could have maintained a positive pressure of at least 1/8-inch water gauge as required by the Technical Specification Surveillance Requirement 3.7.10.4 in accordance with the control room envelope habitability program.

Corrective Actions: The licensees immediate corrective actions included replacing the door seal on October 24 and inspection of the door per their post-maintenance testing program.

Corrective Action References: condition report 202307517

Performance Assessment:

Performance Deficiency: The licensees failure to perform post-maintenance testing on control room door DSK 36043 in accordance with plant procedures was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, control room door DSK 36043 was left in a configuration following corrective maintenance that did not assure operability of the control room envelope during a postulated fuel handling accident. Additionally, it was reasonable to question operability of CREVS due to the degraded door seal and, although operability was ultimately demonstrated, there was a significant loss of margin to operability.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Safety SDP. The finding did not involve potential non-compliance with regulatory requirements for protection of the reactor pressure vessel against fracture; it did not involve fuel bundle misplacement or misorientation in the reactor core; it did not involve low temperature over pressurization (LTOP) or either an inadvertent safety injection actuation or the unavailability of a power operated relief valve or LTOP relief valve or their associated setpoints during LTOP operations when LTOP is required; it did not involve a freeze seal; it did not involve steam generator nozzle dam installation; it did not involve the potential for, or an actual, RCS boron dilution event; it did not degrade the ability to isolate a drain down or leakage path; it did not degrade the ability to close or isolate the containment; it did not degrade the physical integrity of reactor containment; it did not involve a pressurized water reactor ice condenser containment or an AP1000 containment. Therefore, the finding screened as very low safety significance (Green).

Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, licensee personnel failed to fully implement the requirements of the post-maintenance testing procedure which could have detected the problem with the door seal prior to starting core alterations.

Enforcement:

Violation: Technical Specification 5.4.1.a. requires, in part, that written procedures shall be established implemented, and maintained covering the activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33, Appendix A requires, in part, procedures for maintenance that can affect the performance of safety-related equipment. Procedure APA-ZZ-00322, Appendix E, Post-Maintenance Test Program, provides post-maintenance testing requirements following maintenance on safety-related equipment and specifies inspections in accordance with Procedure MPM-SK-QW001, Addendum 5, Service and Inspection of Pressure Doors.

Contrary to the above, on October 13, 2023, the licensee performed corrective maintenance on control room door DSK 36043 by replacing the door seal and did not perform inspections in accordance with procedure MPM-SK-QW001, Addendum 5, as required by procedure APA-ZZ-00322, Appendix E. Specifically, the licensee failed to complete procedure MPM-SK-QW001, Addendum 5, step 5.1.5 which required a visual inspection of the door seal for gaps to ensure the seal would be capable of providing a pressure boundary.

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 71124.01 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: On October 25, 2023, during a routine survey of the maintenance services building, the licensee identified an irradiated excore vessel neutron dosimeters box, with a maximum reading of 10 millirem per hour (mR/hr) at 30 centimeters, not stored within a posted radiation area. The box had been moved from the fuel building truck bay into the maintenance services building on October 24, 2023. Failure to store radioactive material in a posted radiation area is an ineffective radiation protection program barrier designed to protect the worker from unintended exposures to radiation. This failure is classified as a more-than-minor performance deficiency.

Title 10 CFR 20.1003 defines, in part, a radiation area as an area, accessible to individuals, in which radiation levels exceed 5 mR/hr at 30 centimeters from any surface that the radiation penetrates. Title 10 CFR 20.1902(a) stated the licensee shall post each radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution, Radiation Area."

Contrary to the above, on October 24, 2023, a box containing radioactive material with a dose rate exceeding 5 mR/hr (i.e., 10 mR/hr) was not stored within a posted radiation area bearing a conspicuous sign for a radiation area.

The licensee initiated multiple corrective actions in response to the incident: controlled the box once identified and moved it into a correctly posted radiation area, corrected the labeling of the box, lessons learned discussions were conducted in multiple morning meetings, an event review team, a prompt human performance evaluation, and a performance analysis worksheet were performed to determine the probable cause and recommend corrective actions to prevent recurrence.

Significance/Severity: Green. The significance of the finding was assessed using IMC 0609, Appendix C, Occupational Radiation Safety SDP, and because the finding was not:

(1) related to as low as is reasonably achievable planning,

(2) did not involve an overexposure,

(3) did not involve a substantial potential for overexposure, and

(4) the ability to assess dose was not compromised, the finding was determined to be of very low safety significance (Green).

Observation: Negative Trend in Worker Protection Assurance During 71152S Maintenance Activities The inspectors performed a review of potential adverse trends in the licensee's corrective action program with emphasis on worker protection assurance (tagouts) during maintenance activities. The inspectors sampled a population of condition reports during the period of July 1 through December 31, 2023. Based on the samples reviewed, the inspectors identified three instances where the worker protection boundaries were either inadequate for the prescribed maintenance activity, were compromised during work, or failed to provide adequate protection. While degradation was apparent in the implementation of the licensee's worker protection assurance program, the licensee's organization successfully identified these degradations prior to any harmful events. The licensee independently identified this worker protection assurance trend in condition reports 202308879 and 202309050 and is taking appropriate actions to address the degraded conditions that were encountered during the inspection period.

Inadequate Procedure Results in a Fire in the Secondary Alarm Station Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.1] - 71153 NCV 05000483/2023004-03 Resources Open/Closed The inspectors reviewed a self-revealed Green, non-cited violation of License Condition 2.C.(5), Fire Protection Program, regarding an inadequate fire prevention program. Specifically, the licensees fire prevention program did not address overloading electrical outlets which resulted in a fire in an outlet in the secondary alarm station.

Description:

On September 9, 2023, at 4:34 pm, the secondary alarm station (SAS) operator opened fire door 36091 which separates SAS (fire area C-29) from the main control room (fire area C-27) and verbally called away a fire to the control room staff. The source of the fire was an overloaded electrical outlet on the south wall of SAS. The SAS operator saw flames coming out of the outlet and unplugged a surge protector from the outlet that was supplying a portable space heater and a small refrigerator. At 4:42 pm the fire was determined to be out by lack of flame and lowering temperatures on the outlet. The fire brigade was activated, and the control room implemented procedure OTO-KC-00001, Fire Response. The fire brigade responded to the control room, entered SAS, and verified the fire was out before discharging any extinguishing agent. In addition, the breaker feeding the affected electrical outlet was verified to have tripped. Emergency action levels were reviewed by licensed operators who determined that no emergency existed since the fire was extinguished in less than 15 minutes. Control room operators reported an acrid odor in the control room but no smoke or visibility issues. Control Room staff did not have any discomfort from the odor and no spurious actuations of control room equipment occurred. SAS is located within the control room envelope and is the primary egress route to the auxiliary building and the auxiliary shutdown panel in the event of a control room evacuation.

The licensee determined that the electrical outlet had been overloaded. A space heater and a small refrigerator had been plugged into a surge protector being powered from one receptacle of the outlet and a coffee maker was plugged into the second receptacle of the outlet. The inspectors noted a similar event on October 27, 2023, when an electrical outlet in a non-power block building caught on fire due to overloading. During the review of the event, the inspectors were unable to find any licensee procedures addressing electrical outlet loading to ensure the electrical wall outlets were not overloaded.

Corrective Action: The licensees immediate corrective actions included their response to the fire as previously described and an extent-of-condition evaluation of other security stations for similarly overloaded circuits. A communication was also sent plant-wide to emphasize the correct usage of portable electric appliances.

Corrective Action Reference: condition report202306050

Performance Assessment:

Performance Deficiency: The licensees failure to establish adequate programmatic controls to prevent fires due to overloaded electrical outlets was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors 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 and adversely affected the cornerstone objective to limit the likelihood of fires. Specifically, an overloaded outlet resulted in a fire in SAS.

Significance: The inspectors assessed the significance of the finding using AppendixF, Fire Protection Significance Determination Process. The inspectors assigned fire prevention and administrative controls as the fire finding category. Using Appendix F, Attachment 2, Degradation Rating Guidance, dated May 2, 2018, the inspectors assigned a high degradation rating, since it involved findings against the combustible controls program regarding an unapproved heater in the area. This issue was then assigned a fire finding category 1.4.1, Fire Prevention and administrative Controls, with the screening questions determining that the finding was seen as increasing the likelihood of a fire since it resulted in an actual fire and there was no automatic fire detection in SAS, the finding required screening based on the licensees fire probabilistic risk assessment (PRA) results.

According to the fire PRA, SAS is in room 3609 of the control building and is its own fire area, designated C-29, surrounded by a minimum 3-hour fire rated barrier. Area C-29 contains no PRA targets and no cables that support PRA targets in other areas. As a result, this fire area was screened from the fire PRA for individual areas. A multi-compartment analysis was also developed which demonstrated that area C-29 was analyzed as both an exposed compartment and an exposing compartment for all contiguous compartments containing shared penetrations, including the control room. Since there are no PRA targets or cable targets in C-29 it was screened as an exposed compartment. In the analysis performed for C-29 as an exposing compartment it was determined that a hot gas layer cannot be generated in C-29 and, therefore, it screens as an exposing compartment in accordance with the guidance in NUREG/CR-6850. As a result, the final fire PRA model completely screens area C-29 from further consideration. Based on the review completed by a senior reactor analyst, a fire in area C-29 can be considered as a negligible risk impact (i.e., the evaluated delta-CDF was less than 1E-6/year). Therefore, the inspectors determined that the finding had very low safety significance (Green).

Cross-Cutting Aspect: H.1 Resources: Leaders ensure the personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.

Specifically, the licensee did not provide adequate procedure guidance regarding the loading of electrical outlets to prevent fires which resulted in a fire in SAS.

Enforcement:

Violation: License Condition 2.C.(5) requires, in part, that the licensee implement a fire protection program that complies with 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805. Standard NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, section 3.3 requires that a fire prevention program be established, documented, and implemented as part of the fire protection program. Section 3.3.1 requires the fire prevention program to control ignition sources, including programmatic elements necessary to prevent fires from starting.

Contrary to the above, on September 9, 2023, the licensee failed to establish an adequate fire prevention program which would ensure electrical outlets are not overloaded. Specifically, the licensee did not establish adequate programmatic elements, such as a procedure, to minimize the circuit from being overloaded and starting a fire.

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

Failure to Identify and Correct an Inoperable Containment Instrument Sump Level Indicator Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.5] - 71153 Systems NCV 05000483/2023004-04 Operating Open/Closed Experience

The inspectors reviewed a self-revealed Green finding and associated non-cited violation of 10 CFR Part 50 Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, from March 2, through September 30, 2023, the licensee failed to identify and correct an inoperable containment instrument sump level detector. As a result, the containment sump level and flow monitoring system was inoperable for greater than the technical specification allowed outage time. This event was reported as License Event Report 05000483/2023-001-00.

Description:

The containment sump level and flow monitoring system was one of three technical specification leakage detection instrumentation systems that was designed to identify unidentified reactor coolant system leakage by monitoring increases in sump level.

Reactor coolant system (RCS) leakage would result in reactor coolant flowing into the containment normal sumps and into the instrument tunnel sump. Indication of increasing sump level is transmitted to the control room by means of individual sump level transmitters.

This information is used to provide a measurement of leakage by monitoring sump level increase versus time. The containment instrument tunnel sump level indication is summed with level indication from the containment normal sumps to determine a leak rate value for leakage making its way into the containment sumps. The containment sump level monitoring system was designed to detect an RCS leak rate of 0.5 to 1 gpm in accordance with Technical Specification 3.4.15.

To address obsolescence issues associated with an aging containment instrument tunnel sump level detector system the licensee developed modification package MP 20-0012 to install a new level detector system during Refueling Outage 25. The new level detector was installed in May 2022 under job 17005353.

On October 1, 2023, during Refueling Outage 26, per an outage preventive maintenance task, water was added to the containment instrument tunnel sump to verify the performance of the sump pumps and level indicator. Initially, there was no observed change in water level indication. However, about one hour after water addition to the sump had been secured the indicated sump level increased by an 8.3 step change. This prompted the licensee to review recent sump level trend data which indicated a steady water level in the sump of about 5.2 since March 2, 2023. Typically, the licensee would expect a steadily declining sump level due to evaporation. Before the modification was installed, the operators had to add water to the containment sump (approximately once every 3 to 6 months) for the previous sump level detector to indicate a level.

On October 20, 2023, the licensee performed an inspection of the containment instrument tunnel sump level sensing element under job 22003087. The inspection determined that the stem of the level instrument was not centered in the stilling well pipe. This allowed the float to contact the side of the pipe, making it unable to move freely along the stem. Additionally, the stem was found not resting on the bottom of the sump as specified in modification package MP 20-0012.

Based on the above, the licensee determined that the containment instrument sump level detector had stopped responding to changes in sump level as of March 2, 2023, and that it was inoperable since that time. Technical Specification 3.4.15 action (A) requires, in part, that with the containment sump level and flow monitoring system inoperable, within 30 days restore required containment sump level and flow monitoring system to operable status 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 cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Corrective Actions: Immediate corrective actions included lowering the stem of the level detector so that it rests on the sump floor and scheduling a preventative maintenance activity at three-month intervals to add water to the containment instrument tunnel sump to maintain the float above the level where it became stuck.

The licensee is also developing a modification package to install a centering bracket for the float stem and/or remove the stilling well pipe. The modification is planned to be installed during the next refueling outage. The licensee also updated their system trending and monitoring program to look for flatline trends of the containment instrument tunnel sump level detector.

Corrective Action References: condition reports 202306728, 20236800, 20236813, and 202307766

Performance Assessment:

Performance Deficiency: The licensees failure to identify and correct an inoperable containment instrument sump level indicator 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 containment instrument sump level indicator was inoperable from March 2 through September 30, 2023, resulting in a degraded ability to identify RCS leakage.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. In accordance with Section B of IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding is of very low safety significance (Green) because it was a design deficiency that did not represent a loss of operability or PRA functionality.

Cross-Cutting Aspect: P.5 - Operating Experience: The organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee failed to incorporate internal operating experience related to the required frequency of filling the instrument tunnel sump to maintain operability when implementing a new level detector design which led to not identifying the stuck level detector.

Enforcement:

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

Technical Specification 3.4.15, RCS Leakage Detection Instrumentation, requires, in part, that the containment sump level and flow monitoring system shall be operable in Modes 1, 2, 3, and 4. Technical Specification 3.4.15 action (A) requires, in part, that with the containment sump level and flow monitoring system inoperable, within 30 days restore required containment sump level and flow monitoring system to operable status or be in hot standby withing the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Contrary to the above, from March 2 through September 30, 2023, the licensee failed to promptly identify and correct a condition adverse to quality. Specifically, the licensee failed to identify that the instrument tunnel sump level detector was inoperable resulting in its technical specification allowed outage time being exceeded by approximately 6 months. As a result, from March 2 through September 30, 2023, the containment sump level and flow monitoring system was inoperable for greater than the allowed outage time permitted by Technical Specification 3.4.15 action (A), and the unit was not placed in hot standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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 January 9, 2024, the inspectors presented the integrated inspection results to J. Beck, Vice President Nuclear Engineering, and other members of the licensee staff.

• On December 14, 2023, the inspectors presented the inservice inspection results to B. Jungmann, Senior Director, Nuclear Engineering, and other members of the licensee staff.

• On December 4, 2023, the inspectors presented the occupational cornerstone issue follow up inspection results to A. Enloe, Radiation Protection Manager, and other members of the licensee staff.

• On October 10, 2023, the inspectors presented the occupational radiation safety inspection results to K. Scott, Nuclear Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Corrective Action Condition Reports 202307240, 202307241

Documents

71111.01 Procedures OTS-ZZ-00007 Plant Cold Weather 42

71111.04 Drawings M-22EJ01 Residual Heat Removal System 64

71111.04 Procedures OTN-AL-00001, Auxiliary Feedwater System 22

Checklist 1

71111.04 Procedures OTN-EJ-00001 Residual Heat Removal System 28

71111.05 Corrective Action Condition Reports 202305993

Documents

71111.05 Miscellaneous Fire Preplan Manual 41

71111.05 Miscellaneous Fire Safety KC-120, Fire Area C-9

Analysis

Calculation

71111.05 Miscellaneous Fire Safety KC-153, Fire Area RB-1 2

Analysis

Calculation

71111.05 Miscellaneous Fire Safety KC-93, Fire Area A-13

Analysis

Calculation

71111.05 Miscellaneous Fire Safety KC-94, Fire Area A-14

Analysis

Calculation

71111.05 Miscellaneous Fire Safety KC-95, Fire Area A-15

Analysis

Calculation

71111.05 Miscellaneous Fire Safety KC-149 Fire Area D-1

Analysis

Calculation

71111.05 Miscellaneous Jobs 17503769

71111.08P Corrective Action 202205808, 202207467, 202207698, 202207879.

Documents 202208145, 202303051, 202303052, 202303073,

2303961, 202303556, 202306922, 202306957,

2306960, 202307011, 202307060, 202307267,

2307476

71111.08P Corrective Action 202307475, 202307479, 202307481, 202308929

Documents

Resulting from

Inspection

71111.08P Drawings 5042-23-0077 Visual Examination, VT-3 Fuel Pool CLG HX B Aging 10/12/2023

Management Inspection

71111.08P Drawings AB-03-01 S001 Main Steam Loop 3 0

71111.08P Drawings EJ01R017134 Pipe Support Residual Heat Removal System Auxiliary 3

Building "A" Train

71111.08P Drawings ISLT-M-22EJ01 Piping and Instrumentation Diagram Residual Heat Removal H

System

71111.08P Drawings ISLT-M-22EJ01 Piping and Instrumentation Diagram Residual Heat Removal 59

System

71111.08P Drawings S-7.0-58-SA540- Callaway Calibration Notch Fabrication CNGS RPV Stud 0

GRB24-C6.3-CS-Calibration Block WBS: 1-117724

71111.08P Engineering RFR 230281 Accept the Condition of Pressurizer Safety Relief Valve Inlet 0

Evaluations Piping

71111.08P Engineering SCP-PP000-TM-Analytical Assessment of Outer Surface Indications 0

Evaluations PX-000002 Observed on the Inlet Elbow Piping to the Pressurizer

Safety Valves at Callaway

71111.08P Miscellaneous Fourth Interval Code Cases and Requests for Relief/Alternative Applied to 3

Inservice the Fourth Interval

Inspection

Program Plan

Appendix A

71111.08P NDE Reports 1012-005 Ultrasonic Examination, S-BB-02-S001-J Elbow to Flange 04/24/1989

Weld, S-BB-02-S007-J Elbow to Flange Weld

71111.08P NDE Reports 4560-0102 Ultrasonic Examination, 2-BB-02-S001-J Elbow to Flange 12/29/1980

Weld

71111.08P NDE Reports 5000-23-0058 Liquid Penetrant Examination, AE05R027241 12 attachment 10/10/2023

welds

71111.08P NDE Reports 5000-23-0062 Liquid Penetrant Examination, EJ02C022131 8 lugs 10/12/2023

71111.08P NDE Reports 5030-23-001 Radiographic Examination, AE-079-EBB Weld Joint 001 10/13/2023

71111.08P NDE Reports 5030-23-002 Radiographic Examination, AE-079-EBB Weld Joint 002 10/13/2023

71111.08P NDE Reports 5041-23-0044 Visual Examination, VT-2 PM917264 10/16/2023

71111.08P NDE Reports 5042-23-0079 Visual Examination, VT-3 EJ01R017134 10/16/2023

71111.08P NDE Reports 6226-008 Ultrasonic Examination, 2-BB-02-S001-J Elbow to Flange 10/16/1999

Weld

71111.08P NDE Reports 6226-009 Ultrasonic Examination, 2-BB-02-S007-J Elbow to Flange 10/16/1999

Weld

71111.08P NDE Reports UT-023-009 Calibration/Examination Data Sheet: RPV Studs 10/13/2023

71111.08P NDE Reports UT-023-011 Calibration/Examination Data Sheet: Safety Injection 2-EM-10/16/2023

03-S014-C

71111.08P NDE Reports UT-023-012 Calibration/Examination Data Sheet: Main Steam 2-AB-01-10/13/2023

F050

71111.08P NDE Reports UT-023-015 Calibration/Examination Data Sheet: Chemical Volume and 10/16/2023

Control 2-BG-22-F001

71111.08P NDE Reports UT-19-225 Ultrasonic Examination, S-BB-02-S001-J Elbow to Flange 05/23/2019

Weld

71111.08P NDE Reports UT-19-226 Ultrasonic Examination, S-BB-02-S001-J Elbow to Flange 05/23/2019

Weld

71111.08P NDE Reports UT-19-227 Ultrasonic Examination, S-BB-02-S001-J Elbow to Flange 05/23/2019

Weld

71111.08P NDE Reports WDI-PJF-Reactor Vessel Nozzle Inspection Outage RF21 Interval 4 0

1316481-FSR-001 Period 1 2016

71111.08P Procedures 83A1012 Ultrasonic Examination Procedure for Fitting to Fitting Welds 03/13/1989

for Callaway Nuclear Power Plant

71111.08P Procedures 83A6226 Ultrasonic Examination Procedure for Austenitic Piping and 04/12/1997

Vessels =2 Inches Thickness for Callaway Nuclear Power

Plant

71111.08P Procedures APA-ZZ-00661 Administration of Welding 21

71111.08P Procedures APA-ZZ-00661 P-Number Cross Reference Listing 1

Appendix 1

71111.08P Procedures APA-ZZ-00661 WPS Selection Guidelines 2

Appendix 2

71111.08P Procedures APA-ZZ-00661 Personnel Approved to Perform Weld 11

Appendix 3 Inspections/Examinations

71111.08P Procedures APA-ZZ-00662 ASME Section XI Repair/Replacement Program 26

71111.08P Procedures APA-ZZ-00662 ASME Section XI Repair/Replacement Program Mandatory 8

Appendix A Requirements

71111.08P Procedures APA-ZZ-00662 ASME Section XI Code Cases Applied to the Fourth 6

Appendix B Inspection Interval

71111.08P Procedures APA-ZZ-00662 ASME Section XI Repair/Replacement Matrix 5

Appendix E

71111.08P Procedures EDP-ZZ-01003 Inservice Inspection Program 38

71111.08P Procedures EDP-ZZ-01004 Boric Acid Corrosion Control Program 25

71111.08P Procedures LMT-08-PDI-UT-5 Straight Beam Ultrasonic Examination of Bolts and Studs0

71111.08P Procedures LMT-08-UT-PDI-Ultrasonic Examination of Austenitic Piping Welds 0

UT-2

71111.08P Procedures MDP-ZZ-LM001 Fluid Leak Management Program 21

71111.08P Procedures PDI-UT-2 PDI Generic Procedure for the Ultrasonic Examination of 03/11/2020

Austenitic Pipe Welds

71111.08P Procedures QCP-ZZ-05040 Visual Examination to ASME VT-1 26

71111.08P Procedures QCP-ZZ-05041 Visual Examination to ASME VT-2 31

71111.08P Procedures QCP-ZZ-05042 Visual Examination to ASME VT-3 23

71111.08P Procedures QCP-ZZ-05048 Boric Acid Walkdown for RCS Pressure Boundary 15

71111.08P Procedures QCP-ZZ-05049 RPV Head Bare Metal Examination 7

71111.08P Work Orders Job 16504025, 21003312, 22002737

71111.11Q Miscellaneous LOCT 23-5 Reactor Startup 09/01/2023

71111.11Q Miscellaneous LOCT 23-5 Steam Generator Fault 10/25/2023

71111.11Q Procedures OTG-ZZ-00002 Reactor Startup 64

71111.11Q Procedures OTG-ZZ-00003 Plant Startup Hot Zero Power to 30% Power 66

71111.12 Corrective Action Condition Reports 202306579

Documents

71111.12 Miscellaneous Jobs 23502460

71111.13 Miscellaneous Callaway Energy Center RF26 Shutdown Safety 3

Management Plan

71111.13 Procedures APA-ZZ-00315, Risk Management Actions 38

Appendix C

71111.13 Procedures APA-ZZ-00315, Placing Postings for Lowered Inventory

Appendix C,

Checklist 3

71111.13 Procedures APA-ZZ-00315, Configuration Risk Management Program - Shutdown

Appendix E

71111.13 Procedures ODP-ZZ-00002, Protected Equipment Program

Appendix 1

71111.15 Corrective Action Condition Reports 202306728, 202306800, 202306813, 202307060

Documents

71111.15 Miscellaneous Jobs 22503623, 23006728, 23002247

71111.15 Procedures EDP-ZZ-01121 Raw Water Systems Predictive Performance Program 30

71111.20 Miscellaneous RF026 Water Plan

71111.20 Miscellaneous RPV Head Penetrations Examination Report, Callaway 3

Energy Center RF26 Shutdown Safety Management Plan

71111.20 Procedures ETP-BB-03154 Reactor Vessel Head Installation - IPTE 26

71111.20 Procedures OSP-SA-00004 Visual Inspection of Containment for Loose Debris 28

71111.24 Corrective Action Condition Reports 201605144, 201901879, 201903093, 202006446,

Documents 202007059, 202101948, 202202265, 202301933,

2307185, 202307270, 202307305, 202307387,

2307393, 202307517, 202307556, 202307567,

2307842, 202307852, 202307902, 202308492,

2308497

71111.24 Miscellaneous CA 2651 BGHV8160 Past OD 0

71111.24 Miscellaneous CA 2716 Valve Retest Instructions 1

71111.24 Miscellaneous Calculation BG-90 BGHV8152 AOV Capability and Margin Control 0

71111.24 Miscellaneous Jobs 23003300, 18003959, 22002593, 17512666, 21510022,

21001282, 20003577, 23003787, 22513653, 20001021,

23003856, 23004060, 15004072, 21509039, 21502833

71111.24 Miscellaneous MPM-ZZ-QV019 IST Relief Valve Removal and Reinstallation 9

71111.24 Procedures APA-ZZ-00322, Post-Maintenance Test Program 38

Appendix E

71111.24 Procedures APA-ZZ-00750 Hazard Barrier Program 30

71111.24 Procedures ESP-SM-01001 Containment Leakage Rate Testing Program 27

71111.24 Procedures ISF-AE-0L517 S/G A Narrow Range Level Protection 4 COT 21

71111.24 Procedures ISP-BB-0T001 RTD/TC Cross Calibration 16

71111.24 Procedures MSM-KJ-QK001 Emergency Diesel Generator Inspection 45

71111.24 Procedures OSP-BB-VL006 RCS Pressure Isolation Valves Inservice Tests 48

71111.24 Procedures OSP-BG-P005A Centrifugal Charging Pump A Inservice Test - Group B 53

71111.24 Procedures OSP-BG-V0002 Chemical and Volume Control Mode 5 Inservice 24

71111.24 Procedures OSP-EG-LL075 Containment Isolation Valve Leakrate Test 17

71111.24 Procedures OSP-KC-00015 Fire Door Inspections 30

71111.24 Procedures OTN-GK-00001 Control Building HVAC System 68

71111.24 Procedures OTN-GN-00001 Containment Cooling and CRDM Cooling 31

71111.24 Procedures OTN-ZZ-00004 Operation of the Leak Rate Monitor (LRM) 12

71111.24 Procedures OTS-SA-00001 Operation of Engineered Safety Feature Actuation System 30

71111.24 Procedures QCP-ZZ-05041 Visual Examination to ASMI VT-2 31

71114.06 Miscellaneous Emergency

Preparedness

23 ERO

Modified Annual

Drill, August 17 to

December 5, 2023

71124.01 ALARA Plans 16504197, ALARA Post-Work Review - Level 1 Remove/Install 05/04/2022

Revision 2 Insulation for Welds on Pressurizer (TBB03) for UT

Inspections

71124.01 ALARA Plans 22505893-500 Build / Remove Scaffold for Regenerative Heat Exchanger 10/05/2023

Shield Removal

71124.01 ALARA Plans ALARA Package # ALARA Post-Work Review - Level 2 for RWPs R25SGEC 05/18/2022

R25 S/G Primary and R25SGECSETUP Steam Generator Eddy Current

Testing and Setup for all Four Steam Generators

71124.01 ALARA Plans R25REGENHX, ALARA Post-Work Review - Level 1 Install Permanent 05/16/2022

Revision 2 Shielding on Regenerative Heat Exchanger and associated

activities

71124.01 Corrective Action Condition Reports 202202837, 202202907, 202300709, 202303817,

Documents 202305624, 202306025, 202306687, 202307013,

2307061, 202307944

71124.01 Procedures APA-ZZ-00014, Radiation Protection Skill of the Craft 5

Appendix A

71124.01 Procedures APA-ZZ-01001 Callaway Plant ALARA Program 28

71124.01 Procedures APA-ZZ-01004, General Instruction for Donning / Removing Protective 8

Appendix A Clothing

71124.01 Procedures HDP-ZZ-01200 Radiation Work Permits 34

71124.01 Procedures HDP-ZZ-01500 Radiological Postings 53

71124.01 Procedures HDP-ZZ-03000 Radiological Survey Program 52

71124.01 Procedures HDP-ZZ-03000, Frequency and Location of Routine Radiological Surveys 17

Appendix A

71124.01 Procedures HTP-ZZ-02005 Handling of Radioactive Material 52

71124.01 Procedures HTP-ZZ-02023 Unconditional Release of Material from Radiological 26

Controls

71124.01 Procedures HTP-ZZ-06001 High Radiation / Locked High Radiation / Very High 56

Radiation Area Access

71124.01 Procedures HTP-ZZ-06028 Radiological Controls for Pools that Contain or Store Spent 10

Fuel

71124.01 Radiation 2008311307 Air Sample Location: Transfer Canal 08/31/2023

Surveys

71124.01 Radiation 2204290658 Air Sample Location: Reactor Building Steam Generator B/C 04/29/2022

Surveys Primary Side

71124.01 Radiation 2305170029 Air Sample Location: Containment Entry 05/17/2023

Surveys

71124.01 Radiation 2308221715 Air Sample Location: Fuel Building Transfer Canal 08/22/2023

Surveys

71124.01 Radiation 2308281132 Air Sample Location: 2047' Fuel Building Transfer Canal 08/28/2023

Surveys

71124.01 Radiation 2308290850 Air Sample Location: Transfer Canal 08/29/2023

Surveys

71124.01 Radiation CA-M-20220409-RB2047CAV Upper Internals List 04/09/2022

Surveys 17

71124.01 Radiation CA-M-20231004-1 Reactor Building 2047 General Area 10/04/2023

Surveys

71124.01 Radiation CA-M-20231004-Reactor Building 2047 General Area 10/04/2023

Surveys 14

71124.01 Radiation CA-M-20231004-Reactor Building 2047 Refuel Cavity Area 10/04/2023

Surveys 15

71124.01 Radiation CA-M-20231005-RB2023I - Snubber Behind Regen HX Shielding / 10/05/2023

Surveys 16 LHRA/RPB Posting Survey

71124.01 Radiation CA-M-20231006-4 RB2023I - Snubber Behind Regen HX Shielding 10/06/2023

Surveys RB2023I -

Snubber Behind

Regen HX

Shielding

71124.01 Radiation CA-M-20231020- RSS-1 I/S Characterization Survey EVND box 700654 10/20/2023

Surveys 13

71124.01 Radiation CA-M-20231020-9 RSS-1 Characterization Survey shipment 700654 10/20/2023

Surveys

71124.01 Radiation CA-M-20231101-3 RSS-1 Characterization Survey - Container 700654 - 11/01/2023

Surveys Irradiated EVND

71124.01 Radiation Work 22505893500 Remove / Replace Snubbers by Regen Hx 0

Permits (RWPs)

71124.01 Radiation Work R26AOVMOV AOV tests, LimiTorque Service, MOVATS, and associated 0

Permits (RWPs) maintenance

71124.01 Radiation Work R26RCPSEAL Reactor Coolant Pump seal change 0

Permits (RWPs)

71124.01 Radiation Work R26RPCAVITY RP Refuel Cavity job coverage and surveys 0

Permits (RWPs)

71124.01 Radiation Work R26STEC Eddy Current of Flux Thimbles. Flux Thimble Cleaning 0

Permits (RWPs)

71124.01 Radiation Work R26UPPERINLIFT Removal and reinstallation Upper Internals 0

Permits (RWPs)

71124.01 Self-Assessments 202007247-031 Self-Assessment: 05/06/2021

HRA/LHRA Controls

71124.04 Calculations HPCI 0202 Electronic Dosimeter Calibration Adjustment Factor 04/18/2016

71124.04 Calculations HPCI 1602 Reactor Building Neutron Dosimeters Correction Factor 11/30/2020

71124.04 Calculations HPCI 1603 ISFSI Neutron Dosimeters Correction Factor 10/24/2018

71124.04 Corrective Action Condition Reports 202002361; 202002363; 202002364; 202002404;

Documents 202002405; 202002486; 202005007

2007218; 202102992; 202105029; 202202620;

2203687; 202300304; 202304405

71124.04 Miscellaneous Dosimetry 2022 REIRS Data

Records

71124.04 Miscellaneous Dosimetry Dosimetry records associated with facial contamination May 2022

Records

71124.04 Miscellaneous Dosimetry Dosimetry records associated with facial contamination May 2023

Records

71124.04 Miscellaneous Dosimetry Dose Assessment Associated with worker who fell in the April 2022

Records reactor cavity

71124.04 Miscellaneous Dosimetry Form 5s Associated with declared pregnant workers

Records

71124.04 Procedures APA-ZZ-01000 Callaway Energy Center Radiation Protection Program 48

71124.04 Procedures APA-ZZ-01004 General Instructions for Use of Dosimeter of Legal Record4

Appendix D

71124.04 Procedures HDP-ZZ-01200 Radiation Work Permits 34

71124.04 Procedures HDP-ZZ-01300 Internal Dosimetry Program 38

71124.04 Procedures HDP-ZZ-03000 Performing Airborne Radioactivity Surveys 12

Appendix D

71124.04 Procedures HTP-ZZ-01302-Response to Positive Routine or Diagnostic In Vivo Count 22

DTI-Routine/Diag

71124.04 Procedures HTP-ZZ-01433 Personnel Exposure Records 72

71124.04 Procedures HTP-ZZ-04124-Radeco Model Rad Pro 5 Personal Air Sampler Operation 3

DTI-RAD-PRO-5-

OP

71124.04 Procedures HTP-ZZ-06009 Personnel Contamination Assessment and Decontamination 54

71124.04 Procedures RP-DTI-Air Intake Assessment from Air Samples 8

Sample Dose

71124.04 Procedures RP-DTI-EXT-DLR Operations 4

DOS-DLR-

Operations

71124.04 Procedures RP-DTI-In Vitro In-Vitro Bioassay Sample Collection 5

Bioassay

71124.04 Procedures RP-DTI-TRU Alpha Monitoring Facility Characterization 8

Assessment

71124.04 Self-Assessments SA 202000212-NRC Pre-Inspection Self Assessment - Occupational Dose 05/20/2020

013 Assessment

71152S Corrective Action Condition Reports 202306882, 202307253, 202307701, 20238879, 202309050

Documents

71152S Procedures APA-ZZ-00310 Workman's Protection Assurance 76

71153 Corrective Action Condition Reports 202200168, 202203677, 202203737, 202306050,

Documents 202306728, 202306800, 202306813, 202307766,

2307958, 202308090, 202308108, 202308132,

2308204

71153 Drawings M-22EJ01 Residual Heat Removal System 59

71153 Miscellaneous Jobs 22503623, 23006728

71153 Miscellaneous 8375D06 Swing Check Valve Mod 10001CS990000D0 10

71153 Miscellaneous AMNPES00050- Loop Tolerance and Uncertainty Calculation for Instrument 3

CALC-001 Tunnel Sump Level Element LFLE0079 and Containment

Normal Sump East and West Level Transmitters LFLT0089

and LFLT0094

71153 Miscellaneous MP 20-0012-J-Instrument Tunnel Sump Layout Drawing 2

24LF12-002-001

71153 Miscellaneous MP 20-0012-RFR Obsol XMTR-MIN 2

160149

71153 Miscellaneous RFR 230309 TASM - Install Blank Flange in Place of EJF0003 to Support 0

BB8948C Repair

71153 Procedures OSP-ZZ-00001 Control Room Shift and Daily Log Readings and Channel 94 and 95

Checks

71153 Procedures OTN-EJ-00001 Residual Heat Removal System 28

2