IR 05000482/2024002

Integrated Inspection Report 05000482/2024002

ML24200A177
Person / Time
Site:	Wolf Creek
Issue date:	07/18/2024
From:	Greg Werner NRC/RGN-IV/DORS/PBB
To:	Reasoner C Wolf Creek
Werner G
References
IR 2024002
Download: ML24200A177 (1)

Text

July 18, 2024

SUBJECT:

WOLF CREEK GENERATING STATION - INTEGRATED INSPECTION REPORT 05000482/2024002

Dear Cleve Reasoner:

On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Wolf Creek Generating Station. On July 9, 2024, the NRC inspectors discussed the results of this inspection with R. Bayer, General Manager, Plant Operations, and other members of your staff. The results of this inspection are documented in the enclosed report.

Three findings of very low safety significance (Green) are documented in this report. All 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Wolf Creek Generating Station.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC Resident Inspector at Wolf Creek Generating Station. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Gregory E. Werner, Chief Reactor Projects Branch B Division of Operating Reactor Safety Docket No. 05000482 License No. NPF-42

Enclosure:

As stated

Inspection Report

Docket No. 05000482

License No. NPF-42

Report No. 05000482/2024002

Licensee: Wolf Creek Nuclear Operating Corp.

Facility: Wolf Creek Generating Station

Location: Burlington, KS

Inspection Dates: April 1 through June 30, 2024

Inspectors: G. Birkemeier, Resident Inspector N. Greene, Senior Health Physicist C. Henderson, Senior Resident Inspector J. Mejia, Reactor Inspector J. Melfi, Project Engineer D. Proulx, Senior Project Engineer B. Tharakan, Technical Assistant

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 Wolf Creek Generating Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations

Failure to Adequately Barricade and Post a High Radiation Area Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.8] - 71124.01 Radiation Safety NCV 05000482/2024002-01 Procedure Open/Closed Adherence The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of Technical Specification 5.7.1.a, High Radiation Areas, when the licensee failed to adequately barricade and conspicuously post a high radiation area on the 2000-foot elevation of the containment. Specifically, on April 11, 2024, while performing a walk down in the containment, the NRC inspector identified an area outside of a posted and barricaded high radiation area that had accessible dose rates in excess of 100 millirem per hour.

Failure to Follow Licensee Procedures and Implement High Contamination Area Controls Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.11] - 71124.01 Radiation Safety NCV 05000482/2024002-02 Challenge the Open/Closed Unknown The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of Technical Specification 5.4.1.a, Procedures, as recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978, when the licensee failed to adequately implement high contamination area controls for a job activity. Specifically, on April 8, 2024, at least two workers, a welder, and a pipefitter, were exposed to high contamination conditions of 300,000 disintegrations per minute per 100 square centimeters (dpm/100cm2) while performing weld preparation associated with the replacement of train A residual heat removal drain valves EJV0169 and EJV0007 on the pump suction line. As a result, two workers were contaminated and had to be evaluated for unintended internal dose.

Failure to Perform a Radiation Survey to Identify the Magnitude and Extent of Changing Radiological Conditions Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.1] - 71124.01 Radiation Safety NCV 05000482/2024002-03 Resources Open/Closed The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR 20.1501(a) when the licensee failed to survey an area with changing radiological conditions. Specifically, on April 2, 2024, during activities associated with refueling outage 26, the licensee withdrew the incore detector thimbles and failed to survey the area above and around the entrance to the incore tunnel to identify the extent and magnitude of the changing radiation levels. As a result, workers in the high radiation area were not informed about elevated radiation levels as high as 240 millirem per hour at 30 centimeters from the incore tunnel entrance.

Additional Tracking Items

None.

PLANT STATUS

Wolf Creek Generating Station began the inspection period shutdown for refueling outage 26, which began on March 28, 2024. On May 10, 2024, the unit commenced reactor startup, and the reactor was made critical. The unit returned to full power on May 14,

INSPECTION SCOPES

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

71111.01 - Adverse Weather Protection

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 a tornado/missile for the following systems:

• refueling water storage tank on June 12, 2024

• boric acid transfer system on June 12, 2024

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (5 Samples)

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

(1) containment recirculation sump recovery path on April 10, 2024

(2) train A component cooling water system on April 25, 2024

(3) train B offsite circuit to train B safety-related 4160-volt bus (NB02) on April 30, 2024

(4) train B containment coolers on April 30, 2024

(5) train B residual heat removal (RHR) system fire in the control room actions on April 30, 2024

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (7 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) reactor building 2000-foot elevation inside the secondary shield wall, fire area RB-1 and RB-2, on April 2, 2024

(2) reactor building 2026-foot elevation, fire area RB-3, RB-4, RB-7, and RB-8 on April 2, 2024

(3) reactor building 2068-foot elevation, fire area RB-6, on April 16, 2024

(4) reactor building area within primary shield wall, fire area RB-10, on April 16, 2024

(5) reactor building 2047-foot elevation, fire area RB-5 and RB-10, on April 17, 2024

(6) communication corridor fire doors being held open during alternate ventilation lineup, fire area CC-1, on April 23, 2024

(7) south electrical penetration room damaged fireproofing, fire area A-17, on May 7, 2024

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) train A component cooling water heat exchanger performance analysis on April 2, 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 during refuel outage 26, from April 1 to May 23, 2024.

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

• accumulator safety injection, EP-01-MW7152, pipe to valve

• chemical volume control system, BG-02-S056-A, tee to pipe

• chemical volume control system, BG-21-F027A, pipe to pipe

• control rod drive mechanism, CH-213-112-C, nozzle to housing

• control rod drive mechanism, CH-214-112-B, nozzle to housing

• control rod drive mechanism, CH-214-112-G, nozzle to housing

• control rod drive mechanism, CH-215-112-C, nozzle to housing

• main steam, AB-01-F006, elbow to pipe

• main steam, AB-01-F097, pipe to flued head

• reactor coolant pump, PBB01A-Studs, main studs (24)

• reactor coolant system, BB-04-S002-E, reducer to pipe

• reactor coolant system, BB-04-S011-D, reducer to pipe

• reactor coolant system, BB-04-F0001, nozzle to pipe

• reactor coolant system, BB-06-PW6000, pipe to endcap

• RHR, EJ-04-F043, pipe to valve

• RHR, EJ-04-F048A, pipe to valve

• reactor vessel closure head Bolting, CH-STUDS-1 thru 54, closure head studs (54)

• reactor vessel closure head, RBB01, CH-101-104-A, shell to shell

• reactor vessel closure head, RBB01, CH-101-104-B, shell to shell

• reactor vessel closure head, RBB01, CH-101-104-C, shell to shell

• reactor vessel closure head, RBB01, CH-101-104-D, shell to shell

• reactor vessel closure head, RBB01, CH-103-101, dome to shell, one recordable indication being tracked, no change, indication allowable per ASME Sec XI, IWB-3510.

Visual Examination

• essential service water, 1-EF07-C009, piping support

• essential service water, 1-EF01-C002, piping support

• fuel pool cooling, EEC01A-SUP-1, heat exchanger support

• fuel pool cooling, EEC01A-SUP-2, heat exchanger support

• fuel pool cooling, EEC01A-SUP-3, heat exchanger support

• component cooling water, PEG01A-SUP-1, pump support

• reactor vessel closure head, RBB01, bare metal head surface

• reactor vessel internals, upper internals, core barrel, lower internals

Magnetic Particle Examination

• essential service water, BW-4, butt weld, pipe to pipe

• essential service water, WBU-6A, weld build up

• reactor vessel closure head, RBB01, CH-101-101, head to flange weld

Welding Activities

• gas tungsten arc weld, o safety injection system, EMV0066 & EMV0136 valves and piping, W-1A, W-2, W-3A, W-4A, W-5A, W-6A, W-7A o RHR system, EJV0007 & EJV0169 valves and piping, BW-1, BW-2A, BW-3A, BW-4, SW-1A, SW-2, SW-3A 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) Reviewed video of reactor vessel head penetrations 1, 3, 7, 8, 12, 16, 18, 22, 25, 31, 32, 34, 37, 39, 40, 43, 54, 56, 60, 62, 64, 67, 71, 73, 76, 78, and general head area.

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 evaluations and screenings:

• boric acid evaluation, 2024 04-19 BBFT0446

• boric acid evaluation, 2024 04-20 BBFT0017

• boric acid evaluation, 2024 04-20 EMV0057

• boric acid evaluation, 2024 04-21 EPV0001

• boric acid evaluation, 2024 04-28 BGFT0138B

• boric acid evaluation, 2024 04-29 BGHV8146

• boric acid evaluation, 2024 04-30 EJV0179

• Condition report (CR) 10031941 EMV0053

• CR 10031946 BBV0261

• CR 10031947 BBV0241

• CR 10031948 BBPI001

• CR 10031949 BBLT0462

• CR 10031953 BBFI0033

• During walkdown, indications of refueling cavity liner leaks through concrete were observed.

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) A 100 percent full length eddy current inspection of all tubes, with bobbin coil probe, in steam generators A, B, C and D. A total of 45 tubes were plugged, which included 4 tubes in steam generator A, 20 tubes in steam generator B, 12 tubes in steam generator C, and 9 tubes in steam generator D. One tube from steam generator A had an axial indication of outside diameter stress corrosion cracking, located at a dent located at the seventh support plate on the cold leg side, and was plugged.

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 plant startup from refueling outage 26 and low power reactor physics testing on May 10, 2024.

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

(1) The inspectors observed and evaluated just-in-time training for plant startup from refueling outage 26 on April 30, 2024.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (2 Samples)

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

(1) steam generator atmospheric relief valve and turbine-driven auxiliary feedwater pump flow control valve nitrogen accumulator system maintenance effectiveness on May 16, 2024

(2) spent fuel pool cooling maintenance effectiveness on June 10, 2024

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (5 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) emergent activities for the loss of train A safety-related 4160-volt bus during load shedder and emergency load sequencers card replacement on April 8, 2024

(2) turbine-driven auxiliary feedwater pump overspeed failure emergent work activities on May 8, 2024

(3) train B essential service water from component cooler water B heat exchanger isolation valve failed fire protection surveillance testing emergent work activities or May 30, 2024

(4) train A 4kV degraded voltage time delay relay failure emergent work activities on June 12, 2024

(5) ultimate heat sink dredging activities on June 26, 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) train B essential service water piping leak from a weld above drain valve EFV0278 operability determination on April 30, 2024

(2) turbine-driven auxiliary feedwater pump steam generator A flow control valve with a nitrogen leak on the associated atmospheric relief valve operability determination on May 15, 2024

(3) train A safety-related 4160-volt station blackout supply breaker 114 failed to closed during surveillance testing past operability determination on June 11, 2024

(4) train A centrifugal pump coupling grease leakage operability determination on June 11, 2024

(5) emergency diesel generator B number 8-cylinder abnormal low firing pressure and high exhaust temperature operability determination on June 30, 2024

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) startup transformer replacement on April 23, 2024

(2) removal of flux thimble at reactor location L-8 from service on May 15, 2024

71111.20 - Refueling and Other Outage Activities

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

(1) The inspectors evaluated refueling outage 26 from April 1 to May 11, 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) (8 Samples)

(1) emergency diesel generator A maintenance post-maintenance testing on April 18, 2024

(2) containment cooler D motor replacement post-maintenance testing on April 28, 2024

(3) startup transformer replacement post-maintenance testing on April 30, 2024

(4) train A RHR maintenance post-maintenance testing on May 1, 2024

(5) train B RHR maintenance post-maintenance testing on May 1, 2024

(6) train A essential service water pump motor replacement post-maintenance testing on May 1, 2024

(7) emergency diesel generator B maintenance post-maintenance testing on May 15, 2024

(8) train A load shedder and emergency load sequencer card replacement post-maintenance testing on May 15, 2024

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) main steam safety valve lift set point testing on May 6, 2024

(2) engineered safety feature actuation system 18-month surveillance test risk-informed documented evaluation on May 7, 2024

(3) low temperature overpressure protection relief valve testing on May 15, 2024

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

(1) main feedwater bypass regulating valves B and D greater than alert value on June 11, 2024

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

(1) reactor coolant pump B seal water supply valve, BBHV8351B, on May 14, 2024

Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)

(1) reactor coolant system leakage post refueling outage on June 17, 2024

71114.06 - Drill Evaluation

Required Emergency Preparedness Drill (1 Sample)

(1) emergency preparedness drill involving an earthquake causing a loss of offsite power, loss of coolant accident, and a containment leak into auxiliary building on June 18,

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 instructed workers on plant-related radiological hazards and the radiation protection (RP) 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) physical and programmatic controls for highly activated or contaminated materials (non-fuel) stored within spent fuel pool and other storage pools

(2) workers exiting the containment building and performing contamination monitoring during a refueling outage

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) Radiation work permit (RWP) 243220, primary side steam generator eddy current testing (S/G A, B, C, D) and zero entry nozzle dam installation and removal

(2) RWP 243200, RP primary & secondary steam generator rover coverage

(3) RWP 246051, reactor vessel lower internals / core barrel removal and re-installation

(4) RWP 242001, welding activities; work order 19-450006 RHR pump A

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:

(1) very high radiation area - containment building, 2000' in-core tunnel access

(2) locked high radiation area - 2047' containment building reactor head scaffolding

(3) HRA - 2031' radwaste building high integrity container filter cask

(4) HRA - 2000' containment building BCHV-8160l letdown piping

(5) HRA - 2000' containment building inside bio-shield steam generator A and C platforms

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

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

71124.03 - In-Plant Airborne Radioactivity Control and Mitigation

Permanent Ventilation Systems (IP Section 03.01) (2 Samples)

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

(1) control building emergency ventilation system A and B trains (FGK02A and FGK02B)

(2) technical support services emergency ventilation system (VTF-1)

Temporary Ventilation Systems (IP Section 03.02) (2 Samples)

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

(1) high efficiency particulate air (HEPA) filtration systems used on the steam generator A, B, C and D work in containment.

(2) HEPA filtration system and decontamination tent used during steam generator stud cleaning.

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

(1) The inspectors evaluated the licensees use of respiratory protection devices. The NRC credits a baseline sample from the International Atomic Energy Agency Operational Safety Review Team review (under section 7.5.6) conducted March 6-23, 2023.

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.

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

Transportation

Shipment Preparation (IP Section 03.04)

(1) The inspectors observed the preparation of radioactive shipment 24R09 of low specific activity radioactive material on April 10,

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

MS07: High Pressure Injection Systems (IP Section 02.06)===

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

MS10: Cooling Water Support Systems (IP Section 02.09) (1 Sample)

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

BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (1 Sample)

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

BI02: RCS Leak Rate Sample (IP Section 02.11) (1 Sample)

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

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

(1) July 1, 2023, through March 31, 2024 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) July 1, 2023, through March 31, 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) emergency diesel generator B failure to start in manual due to a failed relay and emergency diesel part 21 on May 28, 2024

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

(1) The inspectors reviewed the licensees corrective action program to identify potential trends in post-maintenance testing planning following corrective and planned maintenance activities that might be indicative of a more significant safety issue.

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

(1) emergent activities for the loss of train A safety-related 4160-volt bus during load shedding and emergency load sequencing card replacement on April 23,

INSPECTION RESULTS

Failure to Adequately Barricade and Post a High Radiation Area Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.8] - 71124.01 Radiation Safety NCV 05000482/2024002-01 Procedure Open/Closed Adherence The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of Technical Specification 5.7.1.a, High Radiation Areas, when the licensee failed to adequately barricade and conspicuously post a high radiation area on the 2000-foot elevation of the containment. Specifically, on April 11, 2024, while performing a walk down in the containment, the NRC inspector identified an area outside of a posted and barricaded high radiation area that had accessible dose rates in excess of 100 millirem per hour.

Description:

On April 11, 2024, inspectors walked down the 2000-foot elevation of containment, near the letdown system containment isolation valve. While using a gamma dose rate survey meter, the inspectors identified general area dose rates of greater than 100 millirem per hour outside of the posted HRA boundary. The containment was generally posted as a radiation area. The inspectors noticed a licensee RP technician in the area and asked him to verify the general area dose rates just surveyed outside of the posted HRA boundary. The RP technician came over and surveyed the area with a Ludlum ion chamber survey meter and confirmed that the dose rates were greater than 100 millirem per hour, as observed by the inspectors. The inspectors then proceeded to exit containment for a scheduled meeting.

Upon exiting containment, the inspectors discussed this issue of concern with a RP supervisor waiting outside of containment. The RP supervisor and a RP tech then proceeded to perform additional surveys to evaluate the issue discussed. The licensee found the accessible higher dose rates appeared to stream from a 7-foot-high letdown piping elbow.

However, these post surveys were conducted after the HRA barricade was moved out from the initial boundary location as identified by the inspectors. After further discussion between NRC and the RP staff, it was determined that the HRA initially identified by the inspectors was not adequately posted and barricaded, as required by the licensee's Technical Specification (TS) 5.7.1 and licensee procedure RPP 02-215, Posting of Radiological Controlled Areas," revision 36A, steps 4.15.1, 5.2.1, 5.2.2, 5.3.4, and 9.1.2.

Procedure RPP 02-215, stated the following, in part:

• Step 4.15.1: A High Radiation Area is defined as any area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 0.1 REM in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 centimeters

(cm) from the radiation source or 30 centimeters from any surface that the radiation penetrates.

• Step 5.2.1: Supervisors of RP ensure that plant areas are properly posted.

• Step 5.2.2: Supervisors of RP approve the down posting of high radiation areas and locked high radiation areas.

• Step 5.3.4: RP Technicians obtain approval from RP supervision to down post a high radiation area or a locked high radiation area.

• Step 9.1.2: Posted areas must be clearly and conspicuously marked at all accessible sides and entrances. access points, other than doorways, to either a high radiation area or a locked high radiation area will be conspicuously posted to prevent inadvertent entry into the area.

In this occurrence, the HRA, with dose rates greater than 100 millirem per hour at 30 cm from the source, was not conspicuously posted and adequately barricaded, as required.

Additionally, RP supervision discussed the issue with multiple workers and RP staff and determined that the HRA boundary had been moved or down posted without their approval, as required, prior to the inspectors identifying the HRA dose rates outside of the boundary.

Corrective Actions: Once the HRA dose rates and inadequate posting and boundary were confirmed by RP, RP supervision immediately extended the HRA boundary out approximately 2 feet from its current location, added a steel stanchion, and taped it to the floor to prevent easy removal or movement. These implemented corrective actions established appropriate controls for the HRA posting and barricade. The licensee also coached workers on the requirements to inform RP supervision of moving boundaries, down posting HRAs, and establishing appropriate HRA controls.

Corrective Action References: CRs 10032152 and 10032533

Performance Assessment:

Performance Deficiency: The licensee failed to barricade and conspicuously post an HRA on the 2000-foot elevation of containment.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the program & process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to barricade and conspicuously post an HRA had the potential for workers to inadvertently enter an HRA without knowledge that the dose rates exceeded 100 mrem per hour and accumulating unintended dose.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (i.e., Green) because:

(1) it did not involve as-low-as-reasonably achievable planning or work controls,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, licensee Procedure RPP 02-215 required that HRA boundaries were conspicuously posted and barricaded, that RP supervision ensured that HRAs were posted, and RP supervision approved and was knowledgeable of HRA down postings. However, these requirements were not met when RP supervision determined that the HRA boundary in the 2000-foot elevation of containment had been moved (i.e., down posted) without their knowledge and approval, resulting in an inadequately controlled HRA.

Enforcement:

Violation: Title 10 CFR 20.1901(b), Posting of High Radiation Areas, requires, in part, that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words, CAUTION, HIGH RADIATION AREA, or, DANGER, HIGH RADIATION AREA.

Technical Specification 5.7.1.a, High Radiation Areas, requires, in part, that each entryway to areas with dose rates greater than 100 millirem per hour and not exceeding 1.0 rem per hour at 30 centimeters from the radiation source, or any surface penetrated by the radiation, shall be barricaded and conspicuously posted as an HRA. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

Contrary to the above, on April 11, 2024, the licensee failed to barricade and conspicuously post entryways to a high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION, HIGH RADIATION AREA or DANGER, HIGH RADIATION AREA, as required by 10 CFR 20.1901(b) and TS 5.7.1(a). Specifically, an area in the 2000-foot elevation of containment with radiation levels exceeding 100 millirem per hour at 30 cm from the source but did not exceed 1.0 rem per hour at 30 centimeters from the radiation source, or any surface penetrated by the radiation, was not barricaded or posted as a high radiation area, as required.

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 Licensee Procedures and Implement High Contamination Area Controls Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.11] - 71124.01 Radiation Safety NCV 05000482/2024002-02 Challenge the Open/Closed Unknown The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of Technical Specification 5.4.1.a, Procedures, as recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978, when the licensee failed to adequately implement high contamination area controls for a job activity. Specifically, on April 8, 2024, at least two workers, a welder and a pipefitter, were exposed to high contamination conditions of 300,000 disintegrations per minute per 100 square centimeters (dpm/100cm2)while performing weld preparation associated with the replacement of train A residual heat removal drain valves EJV0169 and EJV0007 on the pump suction line. As a result, two workers were contaminated and had to be evaluated for unintended internal dose.

Description:

After reviewing the assigned radiation work permit (RWP), reviewing survey maps of the work area, and briefly speaking to the welder, the inspectors questioned the licensee about whether the welding activities on RHR piping should have been performed with high contamination area controls in place. On day shift of April 8, 2024, a pipefitter and welder were assigned to grind out a socket weld on the train A RHR suction header drain line containing valves EJV0169 and EJV0007. A HEPA air mover and sampler were already staged in the work area by the night shift workers. The workers were signed onto RWP 24-2001, task 2, Welding Activities. The workers were wearing a full set of protective clothing as required for a contamination area. Intermittent RP coverage was provided.

Upon performing the socket weld, the workers identified that additional RHR piping would need to be accessed due to damaging the piping above the valve EJV0169. Working on this additional piping was a change in their work scope. The pipefitter used a 4-inch grinder to remove the socket weld and a hand-file to dress up the pipe prior to welding. Upon exiting from the radiologically controlled area for the lunch period, the pipefitter was found to have 300 net counts per minute (ncpm) on his beard. He was decontaminated and released. A whole-body count was not immediately performed to assess potential internal exposure at this time. Licensee procedures required a whole-body count if facial contamination is detected. The licensee wrote CR 10031822 for not performing a timely whole-body count.

After lunch, the workers went back to continue with the weld on the RHR piping. The RP technician repositioned the HEPA air mover suction closer to provide better air flow away from the workers. Highly contaminated water was dripping out of the RHR system piping at the rate of one drop per second and was captured on an absorbent pad. The water was interfering with the welding. After performing some welding and spending 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> in the work area, the workers exited the area. Radiation protection personnel surveyed the workers and found contamination on the welder. The welder was found to have 40,000 ncpm (i.e., 400,000 dpm/100cm2 with 10-percent efficiency) on his upper front shirt and 500 ncpm on his face. This time, after the welder was decontaminated, he was immediately sent for a whole-body count. Radiation protection personnel provided the inspectors with initial surveys of the area that indicated contamination levels were as high as 2.4 million dpm/100cm2 beta/gamma and 864 beta/alpha ratio. The pipe was decontaminated to 10,000 dpm/100cm2 beta/gamma before work commenced during the dayshift on April 8th.

A follow up survey, M-20240408-25, was performed in the work area. The welders hood was found to have 30,000 dpm/100cm2 beta/gamma, the water and pipe were 300,000 dpm/100cm2 beta/gamma. The welder stated that water continued to drip out of the pipe during their work. Neither the welder nor the RP technician identified this condition as adverse to the protection provided by a single set of protective clothing nor did they stop work until it was confirmed the accessible contamination levels had increased significantly such that the area, they were working in was a high contamination area, which required additional controls. The inspectors determined this was contrary to the RWP procedure, which required they stop work if radiological conditions change or compliance with the RWP requirements can no longer be achieved. Given that highly contaminated water continued to drip from the pipe while the work was ongoing, the inspectors concluded that the radiological conditions had changed and compliance with the RWP requirements was not achieved.

Corrective Actions: The licensee assessed the internal dose of the contaminated welder and pipefitter and determined neither worker received an internal dose. The licensee conducted a review of their procedures for missed opportunities to prevent future personnel contamination events for similar activities. Radiation protection personnel were provided additional training on compliance with RWP requirements and procedures. Workers were coached on refraining from touching their face or clothing while working in contaminated areas and stopping when faced with uncertain or changing conditions.

Corrective Action References: CRs 10031822, 10031829, and 10032535

Performance Assessment:

Performance Deficiency: The failure to establish high contamination area controls when two workers were commencing job activities in radiological conditions exceeding 100,000 dpm/100cm2 was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the program & process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the licensee's failure to implement high contamination area controls for welding activities on the train A RHR system had the potential to lead to unintended dose for the welder and/or pipefitter and resulted in measured contamination levels as high as 400,000 dpm/100cm2 beta/gamma found on the welder's shirt.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (i.e., Green) because:

(1) it did not involve as-low-as-reasonably achievable planning or work controls,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, the licensee noted that the workers had to change their work scope while accessing additional contaminated piping that contained a constant leakage of highly contaminated water (i.e., surveyed at 300,000 dpm/100cm2), and never stopped the job to determine if the radiological conditions they were exposed to were unknown or of high risk for contamination and possible unintended internal exposure.

Enforcement:

Violation: Technical Specification (TS) 5.4.1.a, states in part, written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978. Regulatory Guide 1.33, section 7.e., Procedures for Control of Radioactivity, recommends in part, adequate procedures for "Radiation Protection." Procedure RPP 02-105, RWP, revision 53, step 5.3.5, requires Stopping work whenever changes in the radiological conditions warrant such action or whenever compliance with the RWP requirements can NO longer be achieved.

Contrary to the above, on April 8, 2024, the licensee failed to follow licensee RP procedures.

Specifically, the licensee failed to follow Procedure RPP 02-105 when they did not stop when changes in the radiological conditions, (i.e., the constant dripping of highly contaminated water from the RHR piping) warranted such action and compliance with RWP requirements could no longer be achieved, by using a single set of protective clothing for a contamination area versus additional protective clothing and equipment for a high contamination condition.

This failure resulted in a personnel contamination event where a welder was found to have high levels of radioactive contamination on his shirt and his face.

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

Failure to Perform a Radiation Survey to Identify the Magnitude and Extent of Changing Radiological Conditions Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.1] - 71124.01 Radiation Safety NCV 05000482/2024002-03 Resources Open/Closed The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR 20.1501(a) when the licensee failed to survey an area with changing radiological conditions. Specifically, on April 2, 2024, during activities associated with refueling outage 26, the licensee withdrew the incore detector thimbles and failed to survey the area above and around the entrance to the incore tunnel to identify the extent and magnitude of the changing radiation levels. As a result, workers in the high radiation area were not informed about elevated radiation levels as high as 240 millirem per hour at 30 centimeters from the incore tunnel entrance.

Description:

On April 9, 2024, during refueling outage (RF) 26, NRC inspectors performed radiation surveys of the Wolf Creek reactor inside the bio-shield on the 2000-foot elevation of containment while the reactor was defueled. The entryway to the bio-shield was posted and barricaded as a high radiation area and contamination area. While performing a survey near the entrance to the incore tunnel, which was locked, enclosed, and posted with signage stating, "Grave Danger - Very High Radiation Area," the inspectors identified dose rates higher than expected based on the high radiation area briefing of radiological conditions within the bioshield and existing radiation protection (RP) survey maps. The inspectors identified that dose rates were over 200 millirem per hour at the ladder access entrance to the incore tunnel. The inspectors informed the licensee's RP staff about the elevated dose rates and the licensee performed confirmatory surveys to validate the inspectors' findings.

The licensee confirmed the dose rates were up to 240 millirem per hour at 30 centimeters from the incore tunnel ladder access and cover plate. These elevated dose rates were not previously recorded by the licensee on survey maps that were being used to brief workers entering the high radiation area. The previous surveys indicated that the highest dose rates in the same area were 12 millirem per hour at 30 centimeters, which were 20 times less than actually encountered. After additional discussions with the licensee, the inspectors determined that after the incore detector thimble tubes were withdrawn on April 2, 2024, no surveys were performed to assess the magnitude and extent of radiation levels in the area.

Subsequent to the withdrawal of the thimble tubes, scheduled refueling outage work activities continued in the area without workers being made knowledgeable of the increased dose rates until April 10th when the licensee performed the confirmatory surveys and updated the applicable survey maps of the area.

The inspectors requested a list of any work activity that had been performed in the area from the time of retracting the thimble tubes on April 2, 2024, until the time of the elevated dose rates discovery on April 9, 2024. The licensee provided information that confirmed work was performed in the area on April 5, 2024, and that workers would not have known the radiation levels had changed such that dose rates were 20 times higher than those briefed to them per the survey at the time.

Corrective Actions: On April 10, 2024, the licensee performed a survey at the incore tunnel entrance. The survey was documented, M-20240411-8, and the status boards were updated to reflect the current radiological conditions with dose rates as high as 240 millirem per hour at 30 centimeters from the incore tunnel entrance. The licensee then evaluated their RP program and process for deficiencies.

Corrective Action References: CRs 10032074 and 10032536

Performance Assessment:

Performance Deficiency: The licensee failed to perform an adequate survey to identify changing radiological conditions after withdrawing the incore thimbles, which resulted in workers not being made knowledgeable of the actual dose rates in the high radiation area that were up to 20 times higher than indicated on survey maps and briefed to the workers.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the program & process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, from April 2-10, 2024, work commenced in the high radiation area above and around the entrance to the incore tunnel without workers being made knowledgeable of the actual dose rates that were up to 20 times higher than indicated on survey maps and briefed to the workers, which challenged the licensee's ability to prevent unplanned dose.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (i.e., Green) because:

(1) it did not involve as-low-as-reasonably achievable planning or work controls,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised.

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

Specifically, the licensee evaluated their program and process and determined that multiple procedures were deficient in information and requirements to evaluate changing radiological conditions. For example, procedures used for the incore thimble work failed to contain specific instructions to survey the area after withdrawal and insertion of the thimbles to ensure that actual radiological conditions were identified and documented to maintain knowledgeable workers in the field.

Enforcement:

Violation: Title 10 CFR 20.1501(a) requires, in part, that each licensee make or cause to be made surveys that

(1) may be necessary for the licensee to comply with the regulations in Part 20 and that surveys

(2) are reasonable under the circumstances to evaluate the magnitude and extent of radiation levels.

Contrary to the above, from April 2-10, 2024, the licensee failed to make or cause to be made surveys that complied with the regulations in Part 20 and were reasonable under the circumstances to evaluate the magnitude and extent of radiation levels in the area above and around the entrance to the incore tunnel. Specifically, the licensee failed to perform surveys after withdrawing the incore detector thimble tubes and evaluate the potential radiological hazards of the changing conditions when dose rates changed from 12 millirem per hour at 30 centimeters to 240 millirem per hour at 30 centimeters, which resulted in inaccurate survey maps being used to brief workers entering the high radiation area and those workers not being made knowledgeable of the actual radiological conditions.

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

Observation: Semiannual Trend 71152S The inspectors reviewed the licensees corrective action program, performance indicators, system health reports, and other documentation available to help identify performance trends that might indicate the existence of a more significant safety issue.

To verify that the licensee was taking corrective actions to address identified adverse trends that might indicate the existence of a more significant safety issue, the inspectors reviewed the corrective action program documentation associated with a potential adverse in performing and planning required post-maintenance testing following maintenance activities with a contributing cause associated with the cross-cutting aspect H.5, Work Management.

Selected examples of these events included:

• NRC Integrated Inspection Report 05000482/2024001 documented non-cited violation (NCV) 2024001-02, Failure to Perform Post-Maintenance Testing Following Maintenance and Installation of the Bypass Reactor Trip Breaker, for the licensees failure to perform appropriate post-maintenance testing following reactor trip breaker maintenance and installation to verify full technical specification operability. Specifically, Work Order (WO) 18-445075-000 failed to provide appropriate instructions for safety-related bypass reactor trip breaker post-maintenance testing following maintenance and installation to verify the breaker was fully operable, by testing the technical specification required P-4 interlock that was later found to be inoperable due to incorrect wiring. The inspectors assigned a cross-cutting aspect of H-5, Work Management, because the licensee inappropriately planned the installation of the vendor overhauled reactor trip breakers in Mode 1 preventing the licensee from performing adequate post-maintenance testing to verify proper operation of the reactor trip breaker in the bypass location to ensure it was fully operable. The licensee-initiated Condition Reports 10031001 and 10031493.

• The inspectors identified a minor finding and associated violation of Technical Specification 5.4.1.a for the licensees failure to implement Procedure AP 16E-002, Post Maintenance Testing, revision 21B, for containment cooler D motor and fan assembly replacement. Specifically, the licensee failed to plan post-maintenance testing to measure containment cooler D air flow capacity following the replacement of the motor and fan assembly. The inspectors determined the cross-cutting aspect H.5, Work Management, was applicable because the licensee failed to implement AP 16E-002 when planning, controlling, and executing work activities associated with the post-maintenance testing development of containment cooler Ds motor and fan assembly replacement during refueling outage 26. The licensee-initiated Condition Report 10032793 and a performed a satisfactory air flow capacity post-maintenance testing.

• The licensee identified a minor finding and associated violation of Technical Specification 5.4.1.a for the licensees failure to perform required post-maintenance testing for auxiliary feedwater flow control valves ALHV0006, ALHV0008, and ALHV0012. Specifically, the licensee failed to perform Procedure STS AL-201C, Turbine Driven Auxiliary Feedwater System Inservice Valve Test, revision 13, in accordance with work orders: WO 53554, WO 53513, and WO 53369. The inspectors determined the cross-cutting aspect H.5, Work Management, was applicable due to the failure to coordinate work activities to address changing priorities across a whole spectrum of activities contributing to nuclear safety. Specifically, when the licensee changed to perform corrective maintenance for ALHV0006, ALHV0008, and ALHV0012 from online to the outage they failed to ensure the post-maintenance testing activities were also changed to the outage. The licensee-initiated Condition Report 10033418 and a performed Procedure STS AL-201C satisfactorily.

• The inspectors reviewed Condition Report 10033122 and the associated performance assessment worksheet for the licensee failing to perform timed stroke testing of the air operated containment purge isolation valve GTHZ0008 following diagnostic testing.

The diagnostic testing procedure provides procedural guidance to remove and re-install the air-lines for GTHZ0008 in accordance with AP 29B-002, ASME Code Testing of Pumps and Valves, revision 12A, and WO 23-486310-002. The diagnostic testing was performed on April 14, 2024, during refueling outage 26 during the defueled window and the station entered Mode 6 on April 24, 2024. The licensees performance assessment worksheet concluded the operator mindset was to consider the need for post-maintenance testing when putting equipment back in service, not when taking it out of service. As a result, the need for the post-maintenance testing was missed. The inspectors reviewed the associated work orders: WO 23-486310-000, WO 23-486310-001, and WO 23-486310-002, and Technical Specification 3.9.4, Containment Penetrations. Technical Specification 3.9.4 states for the containment purge system, each penetration providing direct access from the containment atmosphere to the outside atmosphere either: 1. Closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. Capable of the being closed by an operable containment purge isolation valve. Work Order 23-486310-002 directed the licensee stroke GTHZ0008 in accordance with Procedure STS GP-001, Containment Purge Valves Actuation Test, revision 32. However, the work order was assigned a mode restraint of Mode 4, instead of a mode restraint of Mode 6. From this information the inspectors identified a minor finding and associated violation of Technical Specification 5.4.1.a for the licensees failure to performed required post-maintenance testing for GTHZ0008 prior to entering Mode 6. The inspectors determined the cross-cutting aspect H.5, Work Management, was applicable due to the failure to coordinate work activities to address changing priorities across whole spectrum of activities contributing to nuclear safety. Specifically, the licensee failed to assign the correct mode restraint for WO 23-486310-002 to ensure the post-maintenance testing for GTHZ0008 was completed prior to entering Mode 6 on April 24, 2024.

The inspectors concluded there was an adverse trend in a failure to perform and plan required post-maintenance testing following planned and corrective maintenance with a contributing cause associated with the cross-cutting aspect of H.5 Work Management.

Additionally, the inspectors noted the licensee also identified the adverse trend in a failure to perform and plan required post-maintenance testing. The licensee initiated CRs 10033639 and 10032794 for the adverse trend in post-maintenance testing activities. At the end of the inspection quarter, the licensee was still developing corrective actions to correct the post-maintenance testing adverse trend. The inspectors did not review any cross-cutting themes because none existed at the site.

EXIT MEETINGS AND DEBRIEFS

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

• On April 23, 2024, the inspectors presented the RP inspection results to M. Boyce, Vice President, Engineering, and other members of the licensee staff.

• On May 23, 2024, the inspectors presented the in-service inspection results to M. Boyce, Vice President, Engineering, and other members of the licensee staff.

• On July 9, 2024, the inspectors presented the integrated inspection results to R. Bayer, General Manager, Plant Operations, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Calculations AN 93-065 WCGS IPEEE High Winds 0

and Tornado Evaluation

71111.01 Calculations CWS-SNP- Revised SNP Setpoints 03/01/1984

453C

71111.01 Corrective Condition 2000-1528, 10015262

Action Reports

Documents

71111.01 Corrective Condition 10033969, 10034320

Action Reports

Documents

Resulting from

Inspection

71111.01 Drawings M-12BG03 Piping and Instrumentation 53

Diagram Chemical and

Volume Control System

71111.01 Drawings M-12BG05 Piping and Instrumentation 28

Diagram Chemical and

Volume Control System

71111.01 Miscellaneous M-10BG System Description Chemical 6

and Volume Control System

71111.01 Miscellaneous WCRE-03 Wolf Creek Tank Document 32

71111.01 Miscellaneous WCRE-34 Fourth 10-Year Interval 13

Inservice Testing Basis

Document

71111.01 Procedures AI 14-006 Severe Weather 20

71111.01 Procedures ALR 00-047D RWST Lev LOLO 1 Auto XFR 10B

71111.01 Procedures ALR 00-047E RWST Lev HILO 13

71111.01 Procedures EMG E-0 Reactor Trip or Safety 47

Injection

71111.01 Procedures EMG ES-02 Reactor Trip Response 42

71111.01 Procedures EMG ES-04 EMG ES-04, Natural 24

Circulation Cooldown

71111.01 Procedures OFN BG-009 Emergency Boration 28

71111.01 Procedures OFN SG-003 Natural Events 40

71111.04 Corrective Condition 10031465, 10031524,

Action Reports 10031582, 10031680,

Documents 10031710, 10031715,

10031745, 10031800,

10031854, 10031864,

10031870, 10031882,

10032073, 10032095,

10032099, 10032106,

10032118, 10032183,

10032221, 10032258,

10032275, 10032576,

10032649, 10032772

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Corrective Condition 10031367, 10031952,

Action Reports 10032260

Documents

Resulting from

Inspection

71111.04 Drawings M-12BB01 Piping and Instrumentation 48

Diagram Reactor Coolant

System

71111.04 Drawings M-12BG03 Piping and Instrumentation 53

Diagram Chemical and

Volume Control System

71111.04 Drawings M-12BN01 Piping and Instrumentation 20

Diagram Borated Refueling

Water Storage Tank

71111.04 Drawings M-12EJ01 Piping and Instrumentation 57

Diagram Residual Heat

Removal System

71111.04 Drawings M-12EM01 Piping and Instrumentation 48

Diagram High Pressure

Coolant Injection System

71111.04 Drawings M-12EP01 Piping and Instrumentation 17

Diagram Accumulator Safety

Injection

71111.04 Miscellaneous E-1F9905 Fire Hazard Analysis 14

71111.04 Miscellaneous E-1F9910 Post Fire Safe Shutdown Area 13

Analysis

71111.04 Miscellaneous E-1F9915 Design Basis Document for 16

Procedure OFN RP-017

71111.04 Miscellaneous WCRE-34 Fourth 10-Year Interval 13

Inservice Testing Basis

Document

71111.04 Procedures AP 14A-003 Scaffold Construction 26A

71111.04 Procedures AP 23L-002 Heat Exchanger Program 5

71111.04 Procedures GEN 00-008 RCS Level Less Than Reactor 36

Vessel Flange Operations

71111.04 Procedures OFN BB-031 Shutdown LOCA 37

71111.04 Procedures OFN EJ-015 Loss of RHR Cooling 34

71111.04 Procedures OFN EJ-040 CL Recirc during Mode 3, with 8

Accumulators Isolated, Mode

4, 5, or 6

71111.04 Procedures OFN RP-017 Control Room Evacuation 56

71111.04 Procedures OFN RP-017A Hot Standby to Cold 13

Shutdown form Outside the

Control Room Due to Fire

71111.04 Procedures STN PE-038 Containment Cooler 15

Performance Test

71111.04 Work Orders WO 55242, 55856, 55907, 56780,

56910, 56991, 57658,

Inspection Type Designation Description or Title Revision or

Procedure Date

20-462334-000, 21-475554-

000, 23-483955-000

71111.05 Corrective Condition 10031116, 10031234,

Action Reports 10031523, 10031878,

Documents 10032097

71111.05 Corrective Condition 10032199, 10032254,

Action Reports 10032255

Documents

Resulting from

Inspection

71111.05 Miscellaneous WCRE-35 Boundary Matrix 11

71111.05 Procedures AP 10-103 Fire Protection Impairment 41

Control

71111.05 Procedures AP 10-104 Breach Authorization 40

71111.05 Work Orders WO 56908

71111.07A Calculations EG-08-W Minimum Wall Thickness in 1

Cooling Water Heat

Exchangers EEG01A and

EEG01B

71111.07A Calculations EG-09-W Tube Plugging for CCW Heat 0

Exchangers

71111.07A Corrective Condition 10031516, 10031578,

Action Reports 10031579, 10031581,

Documents 10031582, 10031726,

10031854, 10032260,

10032870

71111.07A Drawings M-1HX001 Heat Exchanger Tube Sheet 82

Map Component Cooling

Water Heat Exchanger "A"

(North End) EEG01A

71111.07A Drawings M-1HX001 Heat Exchanger Tube Sheet 83

Map Component Cooling

Water Heat Exchanger "B"

(South End) EEG01AB

71111.07A Procedures STN PE-033 CCW Heat Exchanger 14

Performance Test

71111.07A Work Orders WO 2023-484909-000 04/11/2024

71111.08P Corrective Condition 10018380, 10018382,

Action Reports 10018387, 10019129,

Documents 10019132, 10019603,

10020507, 10020515,

10020517, 10020493,

10020491, 10020492,

10020507, 10020512,

10020515, 10020517,

10029772, 10030434,

10030796, 10031467,

10031482, 10031486,

Inspection Type Designation Description or Title Revision or

Procedure Date

10031679, 10031706,

10031862, 10031864,

10031896, 10031957,

10031999, 10032067,

10032127, 10032172,

10032298, 10032330,

10032370, 10032379

71111.08P Corrective Condition 10031259, 10031553,

Action Reports 10031924, 10031947,

Documents 10033822, 10033823

Resulting from

Inspection

71111.08P Drawings M-12EM01 Piping & Instrumentation 49

Diagram High Pressure

Coolant Injection System

71111.08P Drawings M-13EJ01 Piping Isometric Residual 17

Heat Removal System

Auxiliary Bldg. "A" Train

71111.08P Drawings M-13EM07 Small Piping Isometric High 2

Pressure Coolant Injection

System - Aux. Bldg.

71111.08P Drawings M-703-00023 SHUPPS Lower Internals 2

Assembly

71111.08P Drawings M-703-00024 SHUPPS Upper Internals 3

Assembly

71111.08P Drawings M-703-00059 4 Loop Lower Radial Support 2

Clevis Insert Gaging and

Assembly

71111.08P Miscellaneous AI 16F-001 Evaluation of Boric Acid 12

Leakage

71111.08P Miscellaneous AI 16F-002 Boric Acid Leakage 13

Management

71111.08P Miscellaneous AP 16A-003 ASME Section XI 20B

Repair/Replacement Program

71111.08P Miscellaneous AP 16F-001 Boric Acid Corrosion Control 11

Program

71111.08P Miscellaneous I-ENG-023 Steam Generator Data 17

Analysis Guidelines

71111.08P Miscellaneous INR-WC- Remote Visual Examination 04/12/2024

1R26-24-002 Indication Notification Report

71111.08P Miscellaneous M-703-00207 Design Specifications for 11/25/2003

W02 Nuclear Reactor Internals

71111.08P Miscellaneous WCRE-30 Inservice Inspection Program 6

Plan Wolf Creek Generation

Station Interval 4

71111.08P Miscellaneous WR 02070-90 Engineering Disposition 04/17/1990

71111.08P NDE Reports 5167 Magnetic Particle 05/07/2024

Examination, WBU-6A

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.08P NDE Reports 5168 Magnetic Particle 05/07/2024

Examination, BW-4

71111.08P NDE Reports STS PE-040A Reactor Coolant System Test 12/8/2022

71111.08P NDE Reports STS PE-040E RPV Head Bare Metal 05/02/2024

Inspection

71111.08P NDE Reports STS PE-040E RPV Head Bare Metal 11/16/2022

Inspection

71111.08P NDE Reports STS PE-040H RPV Head NDE Examination 11/16/2022

71111.08P Procedures 54-ISI-364- Remote Underwater Visual 02/22/2018

009 Examination of Reactor

Pressure Vessels, Vessel

Internals, and Components in

Pressurized Water Reactors in

Accordance with Section XI

71111.08P Procedures 54-ISI-364- Remote Underwater Visual 9

009 Examination of Reactor

Pressure Vessels, Vessel

Internals, and Components in

Pressurized Water Reactors in

Accordance with Section XI.

71111.08P Procedures CWD-1 Control of Welding 14

Documentation

71111.08P Procedures GWS-ASME ASME General Welding 7

Standard

71111.08P Procedures LMT-08-PDI- Ultrasonic Examination of 0

UT-1 Ferritic Piping Welds

71111.08P Procedures LMT-08-PDI- Ultrasonic Examination of 0

UT-2 Austenitic Piping Welds

71111.08P Procedures LMT-08-PDI- Straight Beam Ultrasonic 0

UT-5 Examination of Bolts and

Studs

71111.08P Procedures LMT-08-PDI- Ultrasonic Examination of 0

UT-6 Reactor Pressure Vessel

Closure Head Welds and

Adjacent Base Metal

71111.08P Procedures LMT-21- Manual Phased Array 0

PAUT-028 Ultrasonic Examination of

Pressure Retaining Welds in

Control Rod (CRD) Housings

71111.08P Procedures QCP-20-541 VT-3 Visual Examination 3

71111.08P Procedures STS PE-040E RPV Head Visual Examination 9

71111.08P Work Orders WO 16-418193-000, 17-431456-

001, 19-450006-000,

21-474032-056, 21-474032-

066, 21-474032-067,

2-482436-002, 22-482488-

000, 12003, 12008, 12011,

2013, 12015, 12017, 12346,

Inspection Type Designation Description or Title Revision or

Procedure Date

2348, 55017, 55020

71111.11Q Corrective Condition 10033371

Action Report

Documents

71111.11Q Procedures GEN 00-003 Hot Standby to Minimum Load 111

71111.11Q Procedures RXE 01-002 Reload Low Power Physics 29

Testing

71111.12 Calculations KA-03-W KA System Back-up Nitrogen 4

Accumulators Capacity

Calculation

71111.12 Corrective Condition 33653, 33654, 33758, 121822,

Action Reports 123103, 137915, 142204,

Documents 10007665, 10012448,

10012555, 10030624,

10030630, 10030361,

10030650, 10030667,

10030732, 10030733,

10030734, 10030736,

10030739, 10030769,

10032166, 10032356,

10032400, 10033032

71111.12 Drawings M-11EF01 System Flow Diagram 13

Essential Service Water

71111.12 Drawings M-12EF02 Piping and Instrumentation 46

Diagram Essential Service

Water System

71111.12 Drawings M-12GN01 Piping and Instrumentation 24

Diagram Containment Cooling

System

71111.12 Drawings M-12KA05 Piping and Instrumentation 7

Diagram Compressed Air

System

71111.12 Engineering CCP 07378 Containment Cooler Fusible 0

Changes Link Equivalency Evaluation

71111.12 Engineering DCP 15241 Spent Fuel Pool Alternate 0

Changes Cooling Modification

71111.12 Engineering DRE 20456 AV1 Positioner Gauge Block 0

Changes Removal

71111.12 Miscellaneous M-10GN System Description for 11

Containment Cooling System

71111.12 Miscellaneous M-620 Design Specification for 14

Containment Coolers for Wolf

Creek Nuclear Operating

Corporation (WCNOC)

71111.12 Procedures AP 16-003 Master Lubrication List and 5

Control of Lubricants

71111.12 Procedures AP 23L-002 Heat Exchanger Program 5

71111.12 Procedures AP 26C-004 Operability Determination and 48

Inspection Type Designation Description or Title Revision or

Procedure Date

Functionality Assessment

71111.12 Procedures BD OFN NB- Loss of All AC Power - 3

234 Shutdown Conditions

71111.12 Procedures INC C-0010 7300 Loop P/S Card (NLP1) 3B

71111.12 Procedures INC S-0020 Air Operated Valve (AOV) 6

Diagnostic Testing

71111.12 Procedures OFN EC-046 Fuel Pool Cooling and 10

Cleanup Malfunctions

71111.12 Procedures OFN NB-034 Loss of All AC Power - 35

Shutdown Conditions

71111.12 Procedures STN EF-022B ESW Train B Post-LOCA Flow 12

Balance

71111.12 Procedures STN PE-037A ESW Train A Heat Exchanger 22A

Flow and DP Trending

71111.12 Procedures STN PE-037B ESW Train B Heat Exchanger 25

Flow and DP Trending

71111.12 Procedures STS EF-925A Containment Coolers Flow

Rate (Upon Safety Injection)

Verification - Train A

71111.12 Procedures STS EF-925A Containment Coolers Flow 12

Rate (Upon Safety Injection)

Verification - Train A

71111.12 Procedures STS EF-925B Containment Coolers Flow 13

Rate (Upon Safety Injection)

Verification - Train B

71111.12 Procedures STS KA-010 Accumulator Inservice Check 20

Valve Test

71111.12 Procedures SYS EC-202 Staging Alternate Spent Fuel 4

Pool Cooling

71111.12 Work Orders WO 13326, 16866, 20453, 22872,

50424, 20-459056-000,

17-432437-006, 18-441848-

004, 20-464226-000,

20-467458-000, 21-471451-

000, 21-472154-000,

21-473086-004, 21-473493-

000, 21-474438-000,

2-477036-000, 22-480069-

000, 22-481429-000,

23-483713-000, 23-483892-

000, 23-483899-000,

23-484530-000, 23-488240-

000

71111.13 Corrective Condition 10031603, 10031604,

Action Reports 10033200, 10033869,

Documents 10033938, 10034274

71111.13 Drawings 8025-C- Site Plan 74

KG1202

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.13 Drawings E-13EF04A Schematic Diagram ESW 2

From Component Cooling

Water Heat Exch. ISO Valve

EFHV0060

71111.13 Drawings S-81 Ultimate Heat Sink Plan G

71111.13 Miscellaneous Specification Number C-302 2

71111.13 Procedures OFN NB-030 Loss of AC Emergency Bus 43

NB01 (NB02)

71111.13 Procedures STN FC-002 Aux Feedwater Turbine 34A

Overspeed Test

71111.13 Procedures STN RP-002H ESW TRN B To/From CCW 3

HX B ISO Switches

71111.13 Procedures STS IC-805A Channel Calibration of NB01 15A

Grid Degraded Voltage, Time

Delay Trip

71111.13 Procedures TMP 10-035 Dredging the UHS Channel 2

71111.13 Work Orders WO 17123, 61319, 64522, 66805

71111.15 Calculations KA-03-W KA System Back-up Nitrogen 4

Accumulators Capacity

Calculation

71111.15 Corrective Condition 33653, 33654, 33758,

Action Reports 10005547, 10029678,

Documents 10030624, 10030630,

10030361, 10030650,

10030667, 10030732,

10030733, 10030734,

10030736, 10030739,

10031933, 10032433,

10032656, 10033139,

10033418, 2008-05818

71111.15 Corrective Condition 10034767

Action Report

Documents

Resulting from

Inspection

71111.15 Drawings M-12EF02 Piping and Instrumentation 46

Diagram Essential Service

Water System

71111.15 Drawings M-12KA05 Piping and Instrumentation 7

Diagram Compressed Air

System

71111.15 Miscellaneous M-018-00309 Instruction Manual for W141

Emergency Diesel Generator

System

71111.15 Miscellaneous M-721-00093 Instruction Manual for W50

Charging/Safety Injection

Pump

71111.15 Operability KA-11-001 Operability Evaluation 1

Inspection Type Designation Description or Title Revision or

Procedure Date

Evaluations

71111.15 Procedures AP 26C-004 Operability Determination and 48

Functionality Assessment

71111.15 Procedures CKL KU-120 SBO DG Component 7

Checklist

71111.15 Procedures EMG C-0 Loss of All AC Power 46

71111.15 Procedures QCP-20-502 Magnetic Particle Examination 9

AC/DC Yoke and AC Coil

Techniques

71111.15 Procedures STN KU-001A SBO DG to NB01 Functional 17

Test

71111.15 Procedures STS KA-010 N2 Accumulator Inservice 20

Check Valve Test

71111.15 Procedures STS KU-121 Energizing NB01 from Station 10

Blackout Diesel Generators

71111.15 Work Orders WO 13326, 22872, 40594, 50424,

216, 58244, 08-312917-

005, 20-459655-000, 21-

471249-000, 22-481429-000,

23-484530-000, 23-487665-

000, 23-488240-000

71111.18 Calculations XX-E-006 AC System Analysis 9

71111.18 Corrective Condition 10028675, 10028767,

Action Reports 10029120, 10030207,

Documents 10030607, 10030882,

10031049, 10031320,

10031321, 10031341,

10031502, 10031518,

10031586, 10031664,

10031897, 10031898,

10032140, 10032251,

10032267, 10032341,

10032395, 10032407,

10032444

71111.18 Corrective Condition 10031379, 10034770

Action Reports

Documents

Resulting from

Inspection

71111.18 Drawings M-12BB04 Piping and Instrumentation 31

Diagram Reactor Coolant

System

71111.18 Drawings M-705-00020 Bottom Mounted Instrument W02

Thimble Guide

71111.18 Drawings M-705-00024 Bottom Mounted Instrument W08

Thimble Guide

71111.18 Drawings M-705-00026 Bottom Mounted Instrument W07

Thimble Guide

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.18 Drawings M-705-00029 Bottom Mounted Instrument W01

Thimble Guide

71111.18 Drawings M-764-00316 Bottom Mounted Instrument B2

Thimble Guide

71111.18 Engineering CP 012331 Seal Table Maintenance 0, 1, 2

Changes Fitting Assembly Replacement

71111.18 Engineering DCP 01500 Start-Up Transformer 3

Changes (XMR01) Replacement

71111.18 Engineering DECP 020771 Remove Flux Thimble at 0

Changes Rector Location L-8 from

Service

71111.18 Miscellaneous E-10NF(Q) Load Shedding and 1

Emergency Load Sequencing

71111.18 Miscellaneous M-705-00032 Deep Bore Seal Table W03

Maintenance Fitting Assembly

Installation at Wolf Creek

71111.18 Miscellaneous TR-94-0015 WCNOC Cycle 27 Core 0

Operating Limits Report

(COLR)

71111.18 Miscellaneous WCAP-16676- Analysis Update for the 03/00/2009

NP Inadvertent Loading Event

71111.18 Miscellaneous WCRE-23 Inservice Inspection 2

Classification Basis Document

71111.18 Procedures AI 22D-003 Outage Scope Identification, 13

Control and Risk

Determination

71111.18 Procedures AP 15D-008 Wolf Creek Cyber Security 7

Program

71111.18 Procedures AP 16A-003 ASME Section XI 21

Repair/Replacement Program

71111.18 Procedures AP 16E-002 Post Maintenance Testing 21B

Development

71111.18 Procedures AP 29A-004 ASME Section XI System 12A

Pressure Testing

71111.18 Procedures INC S-0607 Incore Detector Thimble 12

Insertion

71111.18 Procedures STS KJ-001B Integrated D/G and 70

Safeguards Actuation Test -

Train B

71111.18 Procedures STS PE-040A Reactor Coolant System 16A

Pressure Test

71111.18 Work Orders WO 32663, 55830, 57623, 57655,

57757, 58130, 58270,

23-484977-002

71111.20 Corrective Condition 101466, 10006003, 10015377,

Action Reports 10019243, 10028254,

Documents 10030261, 10030278,

10030526, 10030881,

Inspection Type Designation Description or Title Revision or

Procedure Date

10031056, 10031112,

10031126, 10031256,

10031263, 10031266,

10031428, 10031430,

10031467, 10031518,

10031519, 10031537,

10031702, 10031704,

10031760, 10031761,

10031784, 10031813,

10031824, 10031846,

10032104, 10032161,

10033063, 10033065,

10033088, 10033655

71111.20 Corrective Condition 10031377, 10031960,

Action Reports 10033154, 10033155,

Documents 10033156, 10033157,

Resulting from 10034768, 10034769

Inspection

71111.20 Procedures AI 15C-006 Management Oversight 25

Requirements for Infrequently

Performed and Potentially

Degrading Evolutions

71111.20 Procedures AP 14-001 Control of Heavy Loads, 6A

Lifting, and Rigging

71111.20 Procedures AP 16F-001 Boric Acid Corrosion Control 11

Program

71111.20 Procedures FHP 02-007A Reactor Vessel Closure Head 17

Removal

71111.20 Procedures FHP 02-016 Reactor Vessel Lower 15

Internals Removal and

Installation

71111.20 Procedures GEN 00-008 RCS Level Less Than Reactor 36

Vessel Flange Operations

71111.20 Procedures MGM MOOP- Relief Valve Bench Testing 26A

71111.20 Procedures STN EJ-002 Containment Inspection 27

71111.20 Procedures STN PE-040D RCS Pressure Boundary 5

Integrity Walkdown

71111.20 Procedures STS MT-070 ASME Code Testing of 29

Safety/Relief Valves

71111.20 Work Orders WO 33719, 55472, 55705, 60885,

20-461819-000, 20-461819-

001, 20-461819-003, 20-

2020-003, 21-474028-000,

21-474028-001, 21-474028-

003, 21-474101-000,

21-474101-001, 21-474101-

003, 23-487552-000,

Inspection Type Designation Description or Title Revision or

Procedure Date

23-487553-000

71111.24 Corrective Condition 73421, 101466, 10008190,

Action Reports 10008191, 10015380,

Documents 10020251, 10021293,

10028066, 10029827,

10029828, 10030112,

10030189, 10030193,

10030525, 10031127,

10031187, 10031343,

10031400, 10031401,

10031722, 10031805,

10031806, 10031895,

10032019, 10032099,

10032101, 10032207,

10032223, 10032225,

10032226, 10032257,

10032276, 10032408,

10032433, 10032475,

10032477, 10032534,

10032553, 10032570,

10032656, 10032677,

10032730, 10032772,

10032848, 10032963,

10033001, 10033066,

10033139, 10033145,

10033147, 10033149,

10033180, 10033366,

10033857, 10034200

71111.24 Corrective Condition 10031379, 10032272,

Action Reports 10032492, 10032793,

Documents 10033210, 10033213

Resulting from

Inspection

71111.24 Drawings E-025-00007 RHR A to SIS Hot Leg Recirc W14

sheet 236 Loops 2 and 3 ISO

71111.24 Drawings M-12BB03 Piping and Instrumentation 20

Diagram Reactor Coolant

Pump

71111.24 Drawings M-12EF01 Piping and Instrumentation 30

Diagram Essential Service

Water System

71111.24 Drawings M-12EF02 Piping and Instrumentation 46

Diagram Essential Service

Water System

71111.24 Drawings M-12EJ01 Piping and Instrumentation 57

Diagram Residual Heat

Removal System

71111.24 Drawings M-724-00595 Nozzle Type Relief Valve W03

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.24 Drawings M-K2EF01 Piping and Instrumentation 72

Diagram Essential Service

Water System

71111.24 Engineering WCNO#PES- Wolf Creek Unit 1 STRIDE 23- 0

Changes 00069-REPT- 01 Risk Evaluation

2

71111.24 Miscellaneous STRIDE 21-01 Wolf Creek Integrated D/G 0

and Safeguards Actuation

Testing

71111.24 Miscellaneous WCOP-02 Inservice Testing Program for 16

Pumps, Valves, and Snubbers

71111.24 Miscellaneous WCOP-02 Inservice Testing Program for 16

Pumps, Valves, and Snubbers

71111.24 Miscellaneous WCRE-34 Fourth 10-Year Interval 13

Inservice Testing Basis

Document

71111.24 Procedures AP 16E-002 Post Maintenance Testing 21B

Development

71111.24 Procedures AP 21D-004 Control of Containment 11B

Penetrations during Shutdown

Operations

71111.24 Procedures AP 29B-002 ASME Code Testing of Pumps 12A

and Valves

71111.24 Procedures AP 29E-001 Program Plan for Containment 17A

Leakage Measurement

71111.24 Procedures AP 29G-001 RCS Unidentified Leak Rate 6

Monitoring Program

71111.24 Procedures E-13GN02A Schematic Diagrams 1B

Containment Cooler Fans B

and D

71111.24 Procedures GEN 00-006 Hot Standby to Gold 106

Shutdown

71111.24 Procedures MGE LT-100 Llmitorque Valve Operator 13

General Notes and Details

71111.24 Procedures STN EJ-202 RHR System Valve Test 2

71111.24 Procedures STS AE-209A Main Feed Reg Valve Bypass 6

Valve Inservice Valve Test

Train A

71111.24 Procedures STS AE-209B Main Feed Reg Valve Bypass 7

Inservice Valve Test Train B

71111.24 Procedures STS BB-006 RCS Water Inventory Balance 23

Using the NPIS Computer

71111.24 Procedures STS CV-106 Inservice Testing Check Valve 0

Condition Monitoring for

EFV0476, 0478, 0482, 0484

71111.24 Procedures STS GN-001 CTMT Cooling Fans 11

Operability Test

71111.24 Procedures STS KJ-001A Integrated D/G and 73A

Inspection Type Designation Description or Title Revision or

Procedure Date

Safeguards Actuation Test -

Train A

71111.24 Procedures STS KJ-001B Integrated D/G and 70

Safeguards Actuation Test -

Train B

71111.24 Procedures STS KJ-011B DG NE02 24 Hour Run 43

71111.24 Procedures STS MT-008 S/G Spring Loaded Safety 27D

Valve Settings

71111.24 Procedures STS MT-070 ASME Code Testing of 29

Safety/Relief Valves

71111.24 Procedures STS PE-122 LLRT Valve Lineup for 7

Penetration 22

71111.24 Procedures SYS KJ-124 Post Maintenance Run of 76

Emergency Diesel Generator

B

71111.24 Work Orders WO 12362, 12703, 15900, 20864,

23060, 24808, 44349, 46438,

251, 56102, 56910, 57458,

57647, 57695, 59503, 60540,

60580, 60601, 61166, 61169,

14-395986-006, 15-399425-

003, 17-425266-005, 17-

25266-006, 17-425266-007,

17-425266-008, 20-461819-

000, 20-461819-001,

20-461819-003, 20-462020-

003, 21-474028-000,

21-474028-001, 21-474028-

003, 21-474101-000,

21-474101-001, 21-474101-

003, 23-483960-000,

23-485386-000, 23-485745-

000, 23-485746-000,

23-485829-002, 23-485897-

000, 23-487424-000,

23-487552-000, 23-487553-

000, 23-487813-000

71114.06 Corrective Condition 10034406, 10034408,

Action Reports 10034410, 10034413,

Documents 10034417, 10034421,

10034422, 10034426,

10034430, 10034432,

10034433, 10034434,

10034435, 10034437,

10034438, 10034444,

10034447, 10034450,

10034452, 10034454,

10034455, 10034456,

Inspection Type Designation Description or Title Revision or

Procedure Date

10034471, 10034490

71114.06 Corrective Condition 10034519

Action Report

Documents

Resulting from

Inspection

71114.06 Procedures ALR 00-123A MFP B Trip 16

71114.06 Procedures EMG E-0 Reactor Trip or Safety 47

Injection

71114.06 Procedures OFN NB-030 Loss of AC Emergency Bus 43

NB01 (NB02)

71124.01 ALARA Plans RWP 240061 Fuel Building Transfer Canal 10/30/2023

Entry

71124.01 ALARA Plans RWP 244420 Scaffolding Activities 11/01/12023

71124.01 ALARA Plans RWP 246020 Reactor Vessel Head 10/02/2023

Preparation

71124.01 Corrective Condition 10027419, 10027420,

Action Reports 10027133, 10028178,

Documents 10028180, 10028295,

10030172, 10030174,

10030175

71124.01 Corrective Condition 10031817, 10031822,

Action Reports 10031829, 10031843,

Documents 10031852, 10032026,

Resulting from 10032028, 10032034,

Inspection 10032074, 10032152,

10032215, 10032250,

10032533, 10032535,

10032536, 10032537,

10032540

71124.01 Procedures AP 25A-001 Radiation Protection Manual 20A

71124.01 Procedures AP 25A-100 Containment Entry 19A

71124.01 Procedures AP 25A-200 Access to Locked High or 29A

Very High Radiation Areas

71124.01 Procedures AP 25A-401 ALARA Program 28

71124.01 Procedures AP 25A-700 Use of Temporary Shielding 19C

71124.01 Procedures AP 25B-100 Radiation Worker Guidelines 56A

71124.01 Procedures AP 25D-001 Self-Assessment Process 24

71124.01 Procedures AP 26A-009 NRC Interface and Inspection 7

Guidance

71124.01 Procedures INC S-0606 Incore Detector Thimble 9

Withdrawal

71124.01 Procedures RPP 02-105 RWP 53

71124.01 Procedures RPP 02-210 Radiation Survey Methods 51

71124.01 Procedures RPP 02-215 Posting of Radiological 36A

Controlled Areas

71124.01 Procedures RPP 02-405 RCA Access Control 22

71124.01 Procedures RPP 02-510 Discrete Radioactive Particle 10

Inspection Type Designation Description or Title Revision or

Procedure Date

Contamination Control

71124.01 Procedures RPP 02-515 Release of Radioactive 36

Material from the RCA

71124.01 Radiation M-20190919- RadWaste General Floor Area 09/19/2029

Surveys 17 Routine

71124.01 Radiation M-20240330- I/S Bioshield Post-Peroxide 03/30/2024

Surveys 19 Survey B/C Loop down post

71124.01 Radiation M-20240402- Pulling Incore Thimble Tubes 04/01/2024

Surveys 33

71124.01 Radiation M-20240403- Reactor Head Scaffold Survey 04/03/2024

Surveys 29

71124.01 Radiation M-20240404- RadWaste East Title: 04/04/2024

Surveys 13 Thursday Weekly

71124.01 Radiation M-20240404- RB2047C RV Head Stand - 04/04/2024

Surveys 19 Bare Metal Inspection of RV

Head

71124.01 Radiation M-20240407- Socket weld prep work on 04/07/2024

Surveys 24 EJV0169 and EJV0007

71124.01 Radiation M-20240407-5 47' CTMT After Draindown 04/07/2024

Surveys

71124.01 Radiation M-20240407-7 1974 RHR A Drain line 04/07/2024

Surveys removal EJV-Title: 0007, EJV-

0169

71124.01 Radiation M-20240407-8 2000' CTMT Seal Table - 04/07/2024

Surveys Thimble Retraction

71124.01 Radiation M-20240408- Weekly Survey 2000' 04/08/2024

Surveys 10 Containment

71124.01 Radiation M-20240408- Daily 2000' RadWaste 04/08/2024

Surveys 18

71124.01 Radiation M-20240408- PCE#818 04/08/2024

Surveys 35

71124.01 Radiation M-20240408- Contamination survey of "A" 04/08/2024

Surveys 43 RHR Pump Room

71124.01 Radiation M-20240409-2 Weld "A" RHR Pump EJV- 04/08/2024

Surveys 0007 and EJV-0169

71124.01 Radiation M-20240409- RHR "A " Pre-Job Welding 04/09/2024

Surveys 32 Walkdown

71124.01 Radiation M-20240409-9 2026' Containment Weekly 04/09/2024

Surveys Survey

71124.01 Radiation M-20240410- Skimmers labeled with 04/10/2024

Surveys 23 Radioactive Material sticker

71124.01 Radiation M-20240411-8 2000' I/S Bio - Dose Rate 04/10/2024

Surveys Verification at VHRA Incore

Tunnel Boundary

71124.01 Radiation M-20240416- 2000' BCHV-8160 Letdown 04/16/2024

Surveys 30

71124.01 Radiation RWP 230036 Containment access during 0

Work Permits plant modes 1, 2, 3, 4, & 5 for

Inspection Type Designation Description or Title Revision or

Procedure Date

(RWPs) maintenance functions

71124.01 Radiation RWP 230120 LHRA Waste processing 0

Work Permits activities requiring High Risk

(RWPs) Radiation Protection (RP)

Supervisory Oversight

71124.01 Radiation RWP 230610 Prepare Dry Storage 1

Work Permits Container (DSC) and Transfer

(RWPs) Cask (TC) for fuel loading,

load DSC with Fuel, relocate

DSC to Cask Washdown Pit

(CWP), decontaminate

DSC/TC, close DSC, move

DSC/TC to Fuel Building truck

bay

71124.01 Radiation RWP 242001 Welding Activities 0

Work Permits

(RWPs)

71124.01 Self- Audit Report QUALITY ASSURANCE 02/28/2024

Assessments No: 24-01- AUDIT REPORT

RP/PC Radiological Protection and

Process Control

71124.01 Work Orders 23-483995- Perform INC S-0606 In-core 04/02/2024

000 Detector Thimble Withdrawal

71124.03 Corrective Condition 10024319, 10024625,

Action Reports 10025705, 10025717,

Documents 10026482, 10026816,

10027295, 10028605,

10029010, 10030210,

10020326, 10030237

71124.03 Corrective Condition 10030237, 10031852,

Action Reports 10032026, 10032028,

Documents 10032034

Resulting from

Inspection

71124.03 Drawings A-KAT-12 Piping and Instrumentation 6

Diagram: Mechanical HVAC

Plans Sections and Details

71124.03 Drawings M-12GK01 Piping and Instrumentation 14

Diagram: Control Building

HVAC

71124.03 Drawings M-12GK03 Piping and Instrumentation 21

Diagram: Control Building

HVAC

71124.03 Miscellaneous Air Sample Logs: January 1,

23, to March 14, 2024

71124.03 Miscellaneous Updated Safety Analysis 19

Report (USAR), Chapter 9.4 -

Air Conditioning, Heating,

Inspection Type Designation Description or Title Revision or

Procedure Date

Cooling, and Ventilation

71124.03 Miscellaneous USAR, Chapter12.0 - 37

Radiation Protection

71124.03 Miscellaneous System Health Report 04/11/2024

(7/1/2023 - 12/31/2023): Fuel

Building HVAC system

71124.03 Miscellaneous 22-480447- Charcoal Adsorbent In-Place 07/06/2023

000 Leak Test Safety Related

Units - FGK02B

71124.03 Miscellaneous 22-481189- Charcoal Adsorbent Sampling 02/23/2023

000 for Nuclear Safety Related

Units - FGK02A

71124.03 Miscellaneous 22-482265- HEPA Filter In-Place Leak 06/19/2023

000 Test Safety Related Units -

FGK01A

71124.03 Miscellaneous 22-482268- Charcoal Adsorbent Sampling 06/19/2023

000 for Nuclear Safety Related

Units - FGK01A

71124.03 Miscellaneous 23-487062- HEPA Filter In-Place Leak 07/19/2023

000 Test Safety Related Units -

FGK02A

71124.03 Miscellaneous AIF 10-001-02 SCBA Inspections from 02/05/2024

November 4, 2022, to

February 5, 2024

71124.03 Miscellaneous GT 1245302 Self-Contained Breathing 16

Apparatus Training

71124.03 Procedures AI 10-004 Air Fill Station Operation 4A

71124.03 Procedures AI 14-009 Industrial Respiratory 7

Protection Program

71124.03 Procedures AP 25B-600 Respiratory Protection 9

Program at Wolf Creek

71124.03 Procedures RPP 03-210 Internal Exposure Calculations 17D

and Evaluations

71124.03 Procedures RPP 03-310 Respiratory Protection 21

Equipment

71124.03 Procedures RPP 05-201 Eberline RAS-1 Operation for 5

Tritium Air Sampling

71124.03 Procedures RPP 05-205 AMS-4 Operation 11

71124.03 Procedures RPP 05-920 Use of Vacuum Cleaners and 7A

Portable HEPA Ventilation

Units in the RCA

71124.03 Procedures STN PE-004 Charcoal Adsorber In-Place 11

Leak Test Non-Safety Related

Units

71124.03 Procedures STS PE-006A Train A Charcoal Adsorber In- 2

Place Leak Test Safety

Related Units

71124.03 Procedures SYS GK-121 Control Building HVAC 36

Inspection Type Designation Description or Title Revision or

Procedure Date

Operations

71124.03 Procedures SYS GK-123 Control Building A/C Units 37

Startup and Shutdown

71124.03 Procedures SYS VT-100 Technical Support Center 4

HVAC Startup and Shutdown

71124.03 Radiation AS 24-0060 Airborne Radioactivity Record: 02/27/2024

Surveys 2026' Fuel Building

71124.03 Radiation AS 24-0064 Airborne Radioactivity Record: 02/27/2024

Surveys 2026' Fuel Building

71124.03 Radiation M-20240409- S/G B & C Primary Dose Rate 04/09/2024

Surveys 21 & Smear Survey

71124.03 Radiation M-20240409-6 2000' CTMT, S/G "A" & "D" 04/09/2024

Surveys Routine Survey

71124.03 Radiation RWP 243200 Radiation Protection Primary 1

Work Permits and Secondary Steam

(RWPs) Generator Rover Coverage

71124.08 Procedures RPP 07-111 Handling Cartridge Filters 18

71124.08 Procedures RPP 07-123 Preparation and Shipment of 9D

Radioactive Waste and

Material

71124.08 Shipping 24R09 UN2912, Radioactive 04/10/2024

Records materials, low specific activity

(LSA-I), 7, Fissile Excepted,

dry active waste

71151 Procedures DB-EMG ES- Transfer to Cold Leg 18

Recirculation

71151 Procedures EMG ES-12 Transfer to Cold Leg 25

Recirculation

71151 Procedures WCNOC-163 Mitigating System 13

Performance Index (MSPI)

Basis Document

71152A Corrective Condition 108366, 145735, 10019562,

Action Reports 10029246, 10029822,

Documents 10030242, 10031127,

10031663, 10027491

71152A Drawings E-13KJ03A Schematic Diagram Diesel 19

Gen KKJ01B Engine Control

(Start/Stop Circuit)

71152A Drawings E-13KJ03A Schematic Diagram Diesel 17

Gen KKJ01A Engine Control

(Start/Stop Circuit)

71152A Drawings M-12KJ05 Piping and Instrumentation 17

Diagram Standby Diesel

Generator B Intake Exhaust,

F.O., and Start Air Sys.

71152A Engineering CP 20454 Kinectrics Relay/Contact Block 1

Changes Replacement for EDG Room

Panels

Inspection Type Designation Description or Title Revision or

Procedure Date

71152A Miscellaneous M-018-00309 Instruction Manual for W141

Emergency Diesel Generator

System

71152A Procedures CKL ZL-004 Turbine Building Reading 182

Sheet

71152A Procedures STS KJ-015B Manual/Auto Fast Start, Sync 51

and Loading of EDG NE02

71152A Procedures SYS KJ-124 Post Maintenance Run of 76

Emergency Diesel Generator

B

71152A Work Orders WO 26620. 21-469171-006, 21-

469172-000, 56090,

21-469171-013

71152S Corrective Condition 10027129, 10027148,

Action Reports 10031001, 10031493,

Documents 10032794, 10033122,

10033418, 10033639

71152S Drawings M-12GT01 Piping and Instrumentation 38

Diagram Containment Purge

Systems

71152S Miscellaneous M-018-00309 Instruction Manual for W141

Emergency Diesel Generator

System

71152S Procedures AP 16E-002 Post Maintenance Testing 21B

Development

71152S Procedures AP 23AL-002 Heat Exchanger Program 5

71152S Procedures AP 23E-001 Emergency Diesel Generator 8

Reliability Program

71152S Procedures CKL ZL-005A A EDG Operating Log 6A

71152S Work Orders WO 20-461655-004, 23-486310-

000, 23-486310-001,

23-486310-002

71153 Corrective Condition 10031603, 10031604

Action Reports

Documents

71153 Procedures APF 06-002- Emergency Action Levels 2

Technical Bases