IR 05000397/2025003

Integrated Inspection Report 05000397/2025003

ML25350C339
Person / Time
Site:	Columbia
Issue date:	12/17/2025
From:	John Dixon NRC/RGN-IV/DORS/PBD
To:	Schuetz R Energy Northwest
References
IR 2025003
Download: ML25350C339 (0)

Text

December 17, 2025

SUBJECT:

COLUMBIA GENERATING STATION - INTEGRATED INSPECTION REPORT 05000397/2025003

Dear Robert Schuetz:

On September 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Columbia Generating Station. On December 11, 2025, the NRC inspectors discussed the results of this inspection with Jeremy Hauger, acting Site Vice President and other members of your staff. The results of this inspection are documented in the enclosed report.

Due to the temporary cessation of government operations, which commenced on October 1, 2025, the NRC began operating under its Office of Management and Budget-approved plan for operations during a lapse in appropriations. Consistent with that plan, the NRC operated at reduced staffing levels throughout the duration of the shutdown. However, the NRC continued to perform critical health and safety functions and make progress on other high-priority activities associated with the ADVANCE Act and Executive Order 14300. On November 13, 2025, following the passage of a continuing resolution, the NRC resumed normal operations.

However, due to the 43-day lapse in normal operations, the Office of Nuclear Reactor Regulation granted the Regional Offices an extension on the issuance of the calendar year 2025 inspection reports that should have been issued by November 13, 2025, to December 31, 2025. The NRC resumed the routine cycle of issuing inspection reports on November 13, 2025.

Four findings of very low safety significance (Green) are documented in this report. Four 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 Columbia 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 Columbia 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, John L. Dixon, Jr., Chief Reactor Projects Branch D Division of Operating Reactor Safety Docket No. 05000397 License No. NPF-21

Enclosure:

As stated

Inspection Report

Docket Number:

05000397

License Number:

NPF-21

Report Number:

05000397/2025003

Enterprise Identifier:

I-2025-003-0009

Licensee:

Energy Northwest

Facility:

Columbia Generating Station

Location:

Richland, WA

Inspection Dates:

July 1, 2025, to September 30, 2025

Inspectors:

J. Brodlowicz, Resident Inspector

N. Greene, Senior Health Physicist

C. Highley, Senior Resident Inspector

Approved By:

John L. Dixon, Jr., Chief

Reactor Projects Branch D

Division of Operating Reactor Safety

SUMMARY

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

Inappropriate Delay in Entering a Technical Specification Action Statement Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000397/2025003-01 Open/Closed

[H.9] - Training 71111.15 The inspectors identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to follow the operability determination instructions contained in Plant Procedures Manual 1.3.66 (PPM 1.3.66), Operability Determinations, Revision 38, and Maintenance Weld Program 6.1 (MWP-6.1), Weld Repairs of Leaks (Steam and/or Water) in Piping that is in Service, Revision 4. Specifically, the licensee failed to declare the standby service water train B inoperable for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 29 minutes after receiving an inadequate engineering response for additional operability information regarding a through-wall leak in ASME Class 3 piping.

Failure to Complete an Adequate Survey Resulting in Unplanned Uptakes Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025003-02 Open/Closed

[H.8] -

Procedure Adherence 71152A A self-revealed Green finding and associated Non-cited Violation (NCV) of 10 CFR 20.1501(a)was identified when the licensee failed to complete adequate surveys to evaluate the magnitude and extent of radiation levels in a work area. Specifically, two instrumentation and controls technicians were exposed to airborne contamination while tasked to install a new lever switch due to the licensees failure to perform adequate surveys in the work area prior to the job commencing. These two workers received unplanned uptakes of radioactive contaminants.

Failure to Follow the Radiation Work Permit Resulting in Unplanned Uptakes Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025003-03 Open/Closed

[H.14] -

Conservative Bias 71152A A self-revealed Green finding and associated Non-cited Violation (NCV) of Technical Specification 5.4.1.a was identified when the licensee failed to follow procedures in a radiologically controlled area. Specifically, a group of valve technicians performed valve maintenance activities in the drywell and failed to use face shields, as required by their assigned radiation work permit. Additionally, the health physics technician assigned to provide oversight failed to provide adequate coverage on the job, resulting in multiple unplanned uptakes of radioactive contaminants.

Failure to Implement Adequate Corrective Action to Preclude Repetition Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025003-04 Open/Closed

[P.5] -

Operating Experience 71152A The inspectors identified a Green finding and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to implement adequate corrective actions to preclude repetition. Specifically, during multiple uptake events during refueling outage 27, the licensee did not implement adequate corrective actions to preclude repetition associated with White performance issues previously identified during a May 2021 uptake event during refueling outage 25. These performance issues included failure to perform adequate surveys, failure to follow the radiation work permit requirements, failure to use appropriate respiratory protection or face shields, and failure to assign the appropriate risk to work.

Additional Tracking Items

None.

PLANT STATUS

The unit began the inspection period at rated thermal power. On September 20, 2025, the unit was down powered to 70 percent for a rod pattern adjustment. The unit was returned to rated thermal power on September 20, 2025, and remained at or near rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed 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 summer and seasonal high temperatures for the following systems:

• Ashe switchyard, reactor building heating, ventilation, and air conditioning (HVAC), service water, and vital switchgear HVAC on August 6, 2025

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

(1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather due to extreme high temperatures of greater than 100 degrees Fahrenheit on July 9, 2025.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (3 Samples)

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

(1)service water system A lineup following Work Order (WO) 2228318 for flow balance on August 5, 2025 (2)high-pressure core spray service water system lineup during reactor core isolation cooling system outage on September 10, 2025 (3)residual heat removal B system lineup completed on September 12, 2025

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (2 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) TG-1/2 turbine generator building 501-foot elevation on July 17, 2025

(2) RC-3 vertical cable chase all elevations on July 18, 2025

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

(1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill at the turbine generator 441-foot corridor diesel generator area (DGHV-4-1) on August 11, 2025.

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 down power operations in support of 'B' reactor feed pump swap on August 17, 2025.

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

(1) The inspectors observed and evaluated a licensed operator requalification exam on July 22, 2025.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (1 Sample)

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

(1)past trend and frequency of reactor core isolation cooling indicator and switch calibrations on August 18, 2025

71111.13 - Maintenance Risk Assessments and Emergent Work Control

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

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

(1)protection scheme for fuel pool cooling when time-to 200 degrees Fahrenheit was less then 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> on August 18, 2025 (2)low-pressure core spray and residual heat removal A work week protection scheme on September 3, 2025 (3)increased plant fire risk level and implementation of fire risk management actions during reactor core isolation cooling system outage on September 9, 2025

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (3 Samples)

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

(1)high-pressure core spray battery following WO 2228021 on August 5, 2025 (2)250 Volt batteries during single cell charging, Work Request 29184473, on August 21, 2025 (3)service water system B through-wall leak on discharge 3/4-inch instrument line non-isolatable from the pump discharge, Condition Report 00472842 and 00472892, on September 30, 2025

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) (6 Samples)

(1)flex DG-5 run-time maintenance on July 17, 2025

(2) WO 02220819, replacement of adjustable speed drive A runback, coupled run, and trip testing on September 3, 2025

(3) RCIC-P-3 keep fill pump, post-maintenance testing after Engineering Change 19603, replacement of the pump, on September 3, 2025 (4)reactor core isolation cooling pump 1 seal leakage replacement of outboard seal, WO 0217688601, on September 23, 2025 (5)replace difficult to operate reactor core isolation cooling instrument isolation valve, RCIC-V-207, WO 0220953701, on September 23, 2025 (6)replacement of reactor core isolation cooling main steam line C low-pressure relay, RCIC-RLY-K57, WO 0219990001, on September 23, 2025

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) WO 2212794, reactor vessel steam dome high-pressure surveillance on August 13, 2025

(2) ISP-MS-Q928, main steam line high flow channel D channel functional test/calibration, WO 0222486901, on September 16, 2025 (3)pressure testing and operability of the emergency diesel generator 2 fuel oil system, MMP-DO-E002 and OSP-DO-Q702, on September 19, 2025

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

(1) WO 2227836, inservice testing for control room chiller surveillance completed on September 4, 2025

Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)

(1) B5B flow check for beyond design basis flex pump equipment on July 17, 2025

71114.06 - Drill Evaluation

Additional Drill and/or Training Evolution (1 Sample)

The inspectors evaluated:

(1)emergency preparedness drill that exercised the stations ability to respond to an emergency and perform a turnover during the event on September 25,

OTHER ACTIVITIES - BASELINE

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

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

(1) July 1, 2024, through June 30, 2025

MS07: High-Pressure Injection Systems (IP Section 02.06) (1 Sample)

(1) July 1, 2024, through June 30, 2025

MS08: Heat Removal Systems (IP Section 02.07) (1 Sample)

(1) July 1, 2024, through June 30, 2025

MS09: Residual Heat Removal Systems (IP Section 02.08) (1 Sample)

(1) July 1, 2024, through June 30, 2025

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

(1) July 1, 2024, through June 30, 2025

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) The inspectors reviewed the completed and in-progress corrective actions for station tagging events prior to and during refueling outage R27. The number of events and severity was the basis for this selection. Some of the more serious tagging events could have even resulted in personnel injury had the issue not been caught prior to work. The licensee's corrective action program seems to have proven effective in correcting the underlying issues as tagging events have decreased. The inspectors verified the adequateness and effectiveness of these corrective actions, in particular, the operations representative briefing in the work execution center prior to worker signing onto clearance orders.

(2) The inspectors reviewed several condition reports (AR 0470380, 0470395, 0470470, and 0471180), associated with several radioactive contamination uptake events that occurred during the R27 (April - June 2025) refueling and maintenance outage. The inspectors discovered several issues associated with the events violated NRC and licensee requirements, as documented in this report. Additionally, the inspectors identified that the corrective actions to prevent reoccurrence of personnel contamination events that occurred in refueling outage R25 were inadequate.

INSPECTION RESULTS

Inappropriate Delay in Entering a Technical Specification Action Statement Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000397/2025003-01 Open/Closed

[H.9] - Training 71111.15 The inspectors identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to follow the operability determination instructions contained in Plant Procedures Manual 1.3.66 (PPM 1.3.66), Operability Determinations, Revision 38, and Maintenance Weld Program 6.1 (MWP-6.1), Weld Repairs of Leaks (Steam and/or Water) in Piping that is in Service, Revision 4. Specifically, the licensee failed to declare the standby service water train B inoperable for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 29 minutes after receiving an inadequate engineering response for additional operability information regarding a through-wall leak in ASME Class 3 piping.

Description:

On June 22, 2025, at 11:29 p.m., while reactor power was at approximately 20 percent due to the plant being shut down to support the performance of a balance shot on the main generator turbines, Operations identified water on the deck of standby service water B pumphouse. A pinhole leak was discovered on a 3/4-inch instrument line from standby service water pump train B, with an estimated leak rate of 0.0625 gallons per minute (gpm).

The shift manager performed an initial operability determination concluding that the system was operable based on the overall system leakage being less than 5.7 gpm (maximum allowed system leakage), it was not spraying on any other equipment, did not affect the operability of other safety equipment, and the pending additional engineering input, in accordance with the operability Procedure, PPM 1.3.66. However, the nightshift engineers response failed to address ASME Code requirements for through-wall leaks in Class 2 and 3 piping and the shift manager did not question the lack of the ASME Code evaluation. As a result, an incorrect operability determination was made. Following the inspectors inquiry and following a proper structural integrity and flaw evaluation, the determination was that the affected pipe section required replacement. Operations declare the standby service water system train B inoperable.

Corrective Actions: The station declared the standby service water system train B inoperable at 10:14 a.m. on June 23, 2025, and entered Technical Specification (TS) Limiting Condition for Operations (LCO) 3.7.1, condition B. On June 24, 2025, the plant was placed into Mode 4 at 12:01 p.m. which did not require the system to be operable. On June 26, 2025, the section of pipe was replaced at 11:32 a.m. and exited the TS LCO.

Corrective Action References: Condition Reports 00472842 and 00472892

Performance Assessment:

Performance Deficiency: The licensees failure to follow the operability determination process as required by PPM 1.3.66, resulting in a delay in declaring the system inoperable, was within their ability to foresee and correct and should have been prevented. Specifically, not performing the ASME Code and flaw evaluation during the engineering evaluation for additional operability information allowed the service water system train B to not be declared inoperable for an additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 29 minutes in which the pipe could have failed since the system was still in use.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee's failure to follow the operability determination process procedure resulted in a delay of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 29 mins of declaring the B service water system inoperable due to an inadequate response for a request for additional operability information, because it did not address the service water system against the ASME Code requirements concerning the structural integrity and flaw evaluation.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding was determined to be of very low safety significance (Green) because the finding did not:

(1) affect the design or qualification of a mitigating SSC;

(2) represent a loss of probabilistic risk assessments (PRA) function of a single train TS system for greater than its TS allowed outage time;

(3) represent a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time;

(4) represent a loss of the PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;

(5) represent a loss of a PRA system or function as defined in the Plant Risk Information e-Book or the licensees PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and

(6) represent a loss of the PRA function of one or more non-TS trains of equipment designated as risk significant in accordance with the licensees maintenance rule program for greater than 3 days.

Cross-Cutting Aspect: H.9 - Training: The organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. The organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values.

Specifically, the licensee failed to ensure adequate knowledge transfer regarding ASME Code requirements for through-wall leaks in Class 2 and 3 piping. This contributed to the delay in declaring the system inoperable.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that quality-affecting activities be performed in accordance with approved procedures. Plant Procedures Manual 1.3.66 (PPM 1.3.66), "Operability Determinations," Revision 038, a quality related procedure, provides instructions for performing immediate operability determines and request for additional operability information. Specifically, Section 3.2 of PPM 1.3.66 requires, in part, that through-wall leaks in TS ASME Class 2 and 3 components be evaluated for structural integrity using ASME Code methodology in order to determine operability.

Contrary to the above, on June 23, 2025, the licensee failed to evaluate structural integrity using ASME Code methodology in order to determine operability. As a result, the licensee failed to declare the B service water system inoperable for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 29 minutes due to an inadequate engineering evaluation that did not address ASME Code requirements and operations failure to ensure all the requirements for operability met.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Complete an Adequate Survey Resulting in Unplanned Uptakes Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025003-02 Open/Closed

[H.8] -

Procedure Adherence 71152A A self-revealed Green non-cited violation of 10 CFR 20.1501(a) was identified when the licensee failed to complete adequate surveys to evaluate the magnitude and extent of radiation levels in a work area. Specifically, two instrumentation and controls (I&C)technicians were exposed to airborne contamination while tasked to install a new lever switch due to the licensees failure to perform adequate surveys in the work area prior to the job commencing. These two workers received unplanned uptakes of radioactive contaminants.

Description:

On May 9, 2025, during the Unit 1 refueling outage nightshift, radiation workers were briefed to enter the 437-foot level of the radwaste building to perform maintenance activities. The work area location was in the radwaste off gas sump. Specifically, two I&C technicians were tasked to replace a lever switch by cutting out the old switch and connecting a new switch through a conduit and junction box, as two additional workers stood by for support on the platform above the sump. The job required access to the radwaste sump in order to drain piping, cut the old wires, and connect them to the junction box approximately 16 feet in the overhead, at the top of the sump. The workers noted to the NRC inspectors that the junction box was not located as expected.

During discussions with the NRC inspectors, the I&C technicians noted that upon entry to the radwaste sump, they observed there was a powdery substance on the surface of the sump floor which resulted in airborne dust as they traversed the floor to perform the job duties. The licensees health physics (HP) staff that performed the surveys noted in their investigation statements that they believed the floor sediment was hardened, but the floor of the sump was not surveyed. The HP staff also noted there were concerns that the entryway of the off gas sump (i.e., the walls) was contaminated.

The two I&C technicians were signed onto Radiation Work Permit (RWP) 30005194, R27 RCA HRA. The work area was controlled as a high radiation area and a high contamination area and required an HP pre-job briefing per the RWP. The HP pre-job briefing occurred within a short time prior to the job commencing. The briefing covered the RWP requirements, which informed the workers assigned to the job that they would be required to wear double protective clothing to access the sump area. The associated survey noted the maximum contamination level in the work area as 80,000 disintegrations per 100 cm2 (dpm/100cm2).

Although the area was controlled as a high contamination area, the HP staff noted that the contamination levels were not expected to exceed 100,000 dpm/100cm2; thus, the job did not require the Rule of Three. The licensee defines the Rule of Three as one form of engineering control, respiratory protection or a face shield, and air sampling. In this case, the licensee did not implement this rule, although an air sampler attached to a gooseneck was used to monitor the room airborne conditions. Specifically, no respiratory protection or face shields were worn.

The workers noted that the lever switch was located inside the piping, and they had to drain the piping in order to access the switch. Statements made following the event noted that, During the turnover from day shift, they said they were waiting on HP on how they wanted to proceed, it was suspected there was water in the sump. Further, the workers stated, We also mentioned we needed to drain some water from the line, HP recommended capturing the water so we do not turn the floor into mud. In the worker statements, it was noted that the workers inquired about the use of powered air purifying respirators (PAPRs) to perform the duties within the sump, but HP instructed that PAPRs were not necessary because the sump floor was like a hardened riverbed.

Per NRCs review, this aspect of additional contamination from draining pipes and traversing the sump floors was not adequately addressed during the job planning and pre-job brief, and no floor surveys were adequately performed. The HP staff noted to the NRC that they intended to address this issue further prior to the job commencing but was overwhelmed with additional assigned duties. The HP staff did set up an air sampler head into the overhead of the sump, but there was no continuous oversight to monitor the changing radiological conditions. No airborne alarms occurred.

Section 4.5 of licensee Procedure PPM 11.2.13.1, Radiation and Contamination Surveys, Revision 49, states that, An updated survey should be performed prior to allowing workers to enter an area which has not been surveyed within the last 30 days, unless otherwise directed by RP [radiation protection] supervision. Section 5.4.2 of this procedure states, SMEAR enough locations to adequately assess the locations and quantities of surface contamination in the area. Lastly, Sections 5.5.1 of this procedure states, WIPE floors with a cloth mop to obtain a representative sampling of the area; and Section 5.14.7(f) states that, Smear surveys are representative of the area to sufficiently assess general area contamination levels.

As noted above, the contamination levels on the floor of the off-gas sump were not identified prior to the workers entering the sump. Follow-up survey (M20250510-18) identified contamination levels up to 28 millirad per hour per 100 cm2 (millirad/hr/100cm2) for the sump floor. The pre-job survey used for the pre-job brief only showed contamination levels less than 80,000 dpm/100cm2, however no smears were taken on the sump floor. The follow up survey values challenged the licensee's "Rule of Three" requirements considering that 28 millirad/hr/100cm2 would result in greater than 100,000 dpm/100cm2 beta-gamma contamination levels.

Upon exiting, both I&C technicians working within the sump alarmed the ARGOS portal monitors for detectable contamination. After subsequent monitoring, the workers were escorted to the radiologically controlled area egress point to monitor via the personnel contamination monitors. After alarming these monitors, the workers were surveyed, and detectable contamination was found on the face of the technicians. The technicians were then escorted to perform whole body counts, as required by licensee procedures.

Per the initial set of whole body counts provided to the NRC for review, the two workers received radioactivity measurements with an approximate maximum of 564 nanocuries (nCi)of the Cobalt-60 radionuclide. However, the licensee considered this was primarily due to external contamination and did not count this initial activity in their internal dose evaluation provided to the NRC. NRC challenged this approach and requested that the licensee re-evaluate this aspect for a more appropriate internal dose assignment.

The licensee contracted a vendor to conduct a bioassay evaluation, dated July 22, 2025, using urine and feces from the two I&C technicians exposed to radioactive contaminants.

This data was further evaluated by another vendor, documented in an evaluation dated August 27, 2025, which determined lower internal dose values. The greatest contributor to the internal dose was identified as the Cobalt-60 radionuclide. NRC considers the bounding conditions for their assessment of unplanned internal dose to the exposed workers.

Based on the licensees assessment using an inhalation pathway of exposure, the maximum proposed internal doses associated with these uptakes was documented as greater than 10 millirem in unplanned internal dose. The NRC agreed with this but did challenge the licensee on the presence of alpha in one workers bioassay results.

No internal doses were determined as exceeding any regulatory thresholds. This includes any contributions to internal dose from alpha contributing radionuclides.

Corrective Actions:

4. initiated a bioassay dose evaluation for both I&C technicians working in the sump

Corrective Action References: 0470395

Performance Assessment:

Performance Deficiency: The licensee failed to conduct adequate radiological surveys for the off gas sump floor, which resulted in unplanned uptakes and internal dose by two I&C technicians.

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. Additionally, there was a potential this performance deficiency could lead to a more significant radiation safety concern because of an ineffective radiation program barrier. Specifically, a failure to perform adequate radiological surveys of the work area on the sump floor, which was accessible to the workers, resulted in unknown radiological conditions.

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

(1) it was not a finding in ALARA Plans or work controls,

(2) it was not an overexposure,

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

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

Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. The licensees failure to follow steps 4.5, 5.4.2, 5.5.1, and 5.14.7(f) of PPM 11.2.13.1 to conduct adequate pre-job radiological surveys on the sump floor of the work area resulted in the workers exposure to unknown radiological conditions and unplanned internal dose.

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, on May 9, 2025, 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 accessible in the work area. Specifically, the licensee failed to perform surveys on the floors of the off gas sump area, which resulted in inaccurate survey maps being used to plan the work with appropriate radiation protection controls, and adequately brief workers entering the area for work. Thus, those workers were not made knowledgeable of the actual radiological conditions resulting in unplanned internal dose.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Follow the Radiation Work Permit Resulting in Unplanned Uptakes Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025003-03 Open/Closed

[H.14] -

Conservative Bias 71152A A self-revealed Green finding and associated Non-cited Violation (NCV) of Technical Specification 5.4.1.a was identified when the licensee failed to follow procedures in a radiologically controlled area. Specifically, a group of valve technicians performed valve maintenance activities in the drywell and failed to use face shields, as required by their assigned radiation work permit. Additionally, the health physics technician assigned to provide oversight failed to provide adequate coverage on the job, resulting in multiple unplanned uptakes of radioactive contaminants.

Description:

On May 9, 2025, during the Unit 1 refueling outage morning shift, radiation workers were briefed to enter the 548-foot level of the drywell to perform valve maintenance activities. The work area was being controlled as a locked high radiation area and a contamination area. Specifically, two valve technicians were tasked to remove the main steam relief valve (MSRV)-4-B and transfer it onto the body of MSRV-4-C. A third valve technician was tasked to remove the gaskets from MSRV-4-A, MSRV-4-B, and MSRV-4-C in the area prior to replacing the valves onto the body. Thus, all these duties involved breached (i.e., open) MSRVs. There were three additional workers assigned to this job as ancillary support for rigging, but they were not near the work area and were not impacted by the unintended exposures.

All three involved valve technicians were signed onto RWP 30005133, R27 DW MSRV Maintenance LHRA. The work area was controlled as a locked high radiation area and required a health physics (HP) pre-job briefing per the RWP. The HP pre-job briefing occurred within a short time prior to the job commencing. The briefing covered the RWP requirements, which informed the valve technicians assigned to the job that they would be required to wear face shields for the breach of the MSRVs. The briefing also required continuous oversight by a HP technician in the field during the job activity, while the MSRVs were breached.

Upon arriving at the drywell access point, the valve technicians checked in with the field coverage HP technicians and explained their work tasks. They were assigned a field HP technician to provide job coverage, as required. Additionally, the HP technician that facilitated the pre-job briefing spoke with the assigned field HP technician and explained the requirements for the job activity, including wearing face shields for the MSRV breaches and gasket removal and oversight. The assigned field HP technician did not attend the pre-job briefing on the work activity.

Per the RWP and its associated ALARA plan, the licensees Rule of Three was required when the expected local internal or area contamination is greater than 100,000 disintegrations per minute per 100 cm2. In this job, the valve internals, which were breached, were expected to be at least 280,000 disintegrations per minute per 100 cm2 (dpm/100cm2),based on the work package. The Rule of Three requires one form of engineering controls, such as wetting, covering, or the use of a high efficiency particulate air unit; requires respiratory protection or a face shield; and requires air sampling. The licensee confirmed air sampling was in place, but there was no established engineering control or the use of respiratory protection or a face shield. The RWP also stated, HP [health physics] to review job coverage notes prior to covering work on this RWP.

Continuous adequate health physics coverage plays a critical role in radiological safety, primarily ensuring the radiological safety of workers from the hazards on unintended radiation exposure. In this case, the assigned HP technician had performed these oversight duties in at least three previous MSRV maintenance job activities over the last few days, so he was familiar with the job duties and requirements. Upon entry to the drywell to oversee the valve technicians, the field HP technician brought along a couple face shields for the workers to use as needed. The valve technicians also confirmed they had face shields with them.

As the job commenced, the valve technicians rigged the MSRV-4-B and removed it from its seat on the piping, breaching the valve. However, the valve technicians did not wear face shields as these duties occurred, which did not comply with the RWP requirements. This also held true for the valve technician simultaneously removing the gaskets from the breached valve inlets. Although the field HP technician was observing the work, he did not stop the job to instruct the valve technicians to wear the face shields, as required. In fact, as the valve technicians completed removal of the MSRV-4-B from the valve seat on the piping, the field HP technician took surveys of the valve and proceeded to leave the work area to count his contamination smears. The HP technician did not recall providing a stop work order as he left the area and could no longer provide the required HP coverage, which entails a direct line of sight to the workers. Per discussions with the NRC, the assigned field HP technician noted that another field HP technician was replacing him with the job, but the subsequent field HP technician was not aware of this duty as he informed the NRC inspectors during this review.

Thus, the three valve technicians continued their duties without the appropriate oversight or radiological controls in place (i.e., wearing a face shield).

Once the workers completed their duties of breaching the valves, removing the gaskets, and transferring the MSRV-4-B onto MSRV-4-C piping, they left the drywell. Upon exiting, they proceeded to the nearby ARGOS portal monitors to survey themselves for surface contamination and received alarms for detectable contamination. After subsequent monitoring, the workers were escorted to the radiologically controlled area egress point to monitor via the personnel contamination monitors. After alarming these monitors, the workers were surveyed, and detectable contamination was found on the face of two of the valve technicians. The third valve technician had detectable contamination on his shirt. The three valve technicians were then escorted to perform whole body counts, as required by licensee procedures.

Per the initial set of whole body counts provided to the NRC for review, the three workers received radioactivity measurements with an approximate maximum of 307 nanocuries (nCi)of the Cobalt-60 radionuclide, 39 nCi of the Cobalt-58 radionuclide, and 83 nCi of the Zinc-65 radionuclide. The licensee also contracted a vendor to conduct a bioassay evaluation, dated July 22, 2025, using urine and feces from one of these three workers maximally exposed to radioactive contaminants. This data was further evaluated by another vendor, documented in an evaluation dated August 27, 2025, which determined lower internal dose values. The greatest contributor to the internal dose was identified as the Cobalt-60 radionuclide. NRC considers the bounding conditions for their assessment of unplanned internal dose to the exposed workers.

Based on the licensees assessment using an ingestion pathway of exposure, the maximum proposed internal doses associated with these uptakes was documented as less than 10 millirem committed effective dose equivalent (CEDE). However, the NRC challenged the licensees assessment based on the work activity details and the workers bioassay results, as provided. The NRC proposed that the exposure pathway was more of a combination of inhalation and ingestion, resulting in an approximate unplanned internal dose of greater than 10 millirem CEDE.

On May 9, 2025, the licensee completed a follow up survey of the open MSRV internal pipes in the work area, M-20250509-12, which were accessible to the valve technicians upon breaching, and identified that the valve piping inlet smears showed a maximum of 208 millirad per hour per 100 cm2 (millirad/hr/100cm2) beta-gamma, and smears of the valve inlet showed a maximum of 93 millirad/hr/100cm2 beta-gamma. The licensee informed the NRC that based on the specifications of their survey equipment, this valve piping inlet smear would equate to approximately 3,700,000 dpm/100cm2 for beta-gamma contamination. Per the survey, all alpha smears were below 20 dpm/100cm2.

The licensees position is that the valve technicians access to this level of internal piping contamination was not considered because the workers were not instructed to stand at the plane of the breached valve as it was removed, but they were required to wear face shields for radiological protection. The licensee acknowledged they failed to consider the full radiological impact of removing the gaskets from the piping. The licensee believed that the workers were contaminated primarily by the gasket removals which likely released some internal contamination from the valve inlet. It is noteworthy to mention that the contamination was reportedly present around the workers nose and mouth areas.

The licensee also performed a total effective dose equivalent - as low as reasonably achievable (TEDE-ALARA) evaluation, dated May 10, 2025, which did not use accurate beta-gamma smearable contamination data based on survey M-20250510-6. The TEDE-ALARA evaluation used a contamination value of 500,000 dpm/100cm2, whereas the maximum beta-gamma levels shown on the M-20250510-6 survey was 200 millirad/hr/100cm2. This maximum beta-gamma contamination data would yield a difference of approximately seven times greater in radiological impacts. However, the results still required the use of respiratory protection for this work.

In reviewing the licensee's Procedure PPM 1.3.76, "Integrated Risk Management," the NRC determined that the work activity should have been screened and assessed as a "high" radiological risk activity. This was based on answering question 8 to Attachment 9.10, "Radiological Risk Assessment Checklist," in PPM 1.3.76. Question 8 inquired if the "Activity involves flushing, draining, venting, or breaching of contaminated systems and has the potential to cause spread of contamination to a non-contaminated area, and contact dose rates are greater than or equal to 1000 mrem/hr." NRC determined this response was 'YES' because the maintenance activity involved breaching a highly contaminated system (the MSRV valve piping and inlet), where survey M-20250509-12, dated May 9, 2025, shows a maximum contact dose rate for a valve inlet piping as 5400 millirad/hr, and a maximum contact dose rate of 4040 millirad/hr for a valve inlet. Due to inadequate radiation protection (RP) controls, personnel contamination events occurred which moved contamination to non-contaminated areas of the plant while on the noted workers.

Based on the inspectors review of all work details and discussions with the involved workers, the NRC determined that although the accessible contamination levels to the valve inlets were in excess of 200 millirad/hr/100cm2 beta-gamma, the workers completed the job duties and the bioassay data provided did not support any significant data, for either beta-gamma or alpha emitting radionuclides, that would yield a substantial potential for any overexposures.

No internal doses were determined as exceeding any regulatory thresholds.

Corrective Actions:

4. initiated a bioassay dose evaluation for the valve technician with the maximum

exposure to radioactive contaminants Corrective Action References: 0470380

Performance Assessment:

Performance Deficiency: The licensee failed to follow radiation work permit requirements which resulted in unplanned uptakes and internal dose by valve technicians.

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. Additionally, there was a potential this performance deficiency could lead to a more significant radiation safety concern because of an ineffective radiation program barrier. Specifically, errors related to radiological safety were made by the assigned field HP technician providing job oversight that failed to stop work when the valve technicians deviated from the RWP and associated ALARA plan by not wearing face shields during a MSRV breach and gasket removals.

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

(1) it was not a finding in ALARA Plans or work controls,

(2) it was not an overexposure,

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

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

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. The failure to follow the RWP requirements and proceed with the job activities without adequate consideration of the radiological impacts by not wearing face shields resulted in multiple uptakes and unplanned internal dose.

Enforcement:

Violation: Technical Specification 5.4.1.a states, in part, that licensees shall implement procedures found in Regulatory Guide 1.33, Appendix A, February 1978. Section 7.e.(1) of this appendix describes procedures for Access Control to Radiation Areas Including a RWP System.

The licensee established RWP 30005133, R27 DW MSRV Maintenance LHRA for maintenance work on MSRV-4-A, MSRV-4-B, and MSRV-4-C. The RWP stated, in part, A FACESHIELD IS REQUIRED FOR VALVE BREACH AND INSULATION REMOVAL and CONTINUOUS HP COVERAGE IS REQUIRED FOR MSRV BREACH.

Contrary to this requirement, on May 9, 2025, the licensee failed to implement the requirements of RWP 30005133. Specifically, the workers involved in the valve breach and gasket removals failed to wear face shields, as required by the RWP and job instructions.

Additionally, the assigned field HP technician failed to provide continuous coverage during the job activity as the MSRVs were breached. These failures to follow the RWP requirements, and establish appropriate radiation protection controls, resulted in uptakes and unplanned internal dose.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Implement Adequate Corrective Action to Preclude Repetition Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025003-04 Open/Closed

[P.5] -

Operating Experience 71152A The inspectors identified a Green finding and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to implement adequate corrective actions to preclude repetition. Specifically, during multiple uptake events during refueling outage 27, the licensee did not implement adequate corrective actions to preclude repetition associated with White performance issues previously identified during a May 2021 uptake event during refueling outage 25. These performance issues included failure to perform adequate surveys, failure to follow the radiation work permit requirements, failure to use appropriate respiratory protection or face shields, and failure to assign the appropriate risk to work.

Description:

On May 28, 2021, the licensee had multiple performance deficiencies that resulted in a contamination uptake event related to pipefitters working on the reactor water cleanup (RWCU) heat exchanger (HX) during their refueling outage 25. In response, NRC Inspection Report 05000397/2023090, dated June 1, 2023 (ML23111A2337), identified a White performance issue at Columbia Generating Station, which documented a White finding and three associated violations:

(1) 10 CFR 20.1701 for failure to use adequate engineering controls,

(2) TS 5.7.2 for failure to control access to, and activities in, a high radiation area with dose rates greater than 1.0 rem per hour at 30 centimeters, and

(3) 10 CFR 20.1501(a)(2) for failure to perform adequate surveys. In response to this White finding, the NRC performed a follow up inspection per Inspection Procedure (IP) 95001 and documented the results of the inspection in NRC Inspection Report 05000397/2024040, dated April 2, 2025 (ML25073A062).

The IP 95001 report documented that relative to the May 2021 contamination uptake event, the licensee identified that root and contributing causes of the White finding were related to the use of inadequate RP engineering controls, RWPs, and surveys. In the NRCs review, the inspectors determined that the licensee needed to revise their root cause evaluation (RCE)regarding the identification of their root causes and contributing causes. This updated RCE indicated that the application of rigorous job planning was a root cause and required corrective actions to preclude recurrence (CAPR). Specifically, it was documented in the licensees updated RCE that during the pre-outage planning for the RWCU HX job during refueling outage 25, the pipe weld preparation activity was categorized as a Medium Elevated Risk instead of High Risk. This failure led to less rigorous radiation protection controls and multiple contamination uptakes for the workers.

As documented in the licensees updated RCE for the event, two CAPRs were required to effectively address the causes related to the first White finding:

(1) the conduct of dynamic learning activities to evaluate RP technicians and supervisors to include fundamental radiation standards; positive RP command and control of radiological work activities; decisions related to control of radiological jobs are prudent over simply allowable approach; and

(2) update plant Procedure PPM 11.2.2.12, Radiological Risk Assessment and Management, to define risk mitigation and risk elimination actions; determine initial risk assuming no elimination or mitigation actions; and to only allow the risk categorization to credit the elimination of risk; as well as update Form 26840 to assess a job for high risk work activities.

With these CAPRs in place, future contamination uptake events should have been extremely limited. However, during the licensees most recent refueling outage (R27), the NRC noted at least four different events, within a short time period, that resulted in uptakes of radioactive contaminants by workers. Of these four events, three had related causes to the May 2021 contamination uptake event, and two of the four resulted in unplanned internal doses exceeding 10 millirem committed effective dose equivalent (CEDE).

The licensee documented these events in the following action requests (AR):

AR 470380: This contamination event occurred on May 9, 2025, when three workers received uptakes while performing work on the MSRV gaskets. Specifically, the workers were breaching the MSRVs and were not wearing face shields as required by RWP 30005133, R27 DW MSRV Maintenance LHRA. Nor did the RP technician providing oversight require the workers to wear face shields or provide continuous coverage throughout the MSRV tasks.

The work activity of breaching the highly contaminated systems and accessible survey levels identified on Survey M-20250509-12 supported a "high risk" job RWP and the Rule of Three, but the RWP was noted as "medium risk," and adequate RP controls were not implemented. This resulted in at least one uptake for a valve technician exceeding 10 millirem CEDE. It was determined this issue was primarily attributed to a failure to follow the RWP requirements to wear face shields and provide continuous RP coverage. This was documented as a violation of licensee requirements in which additional details are provided.

AR 470395: This contamination event occurred on May 9, 2025, when two workers received uptakes while performing maintenance activities on the 437-foot level of the radwaste building. The precise work area location was in the radwaste off gas sump. Specifically, two I&C technicians were tasked to replace a lever switch by cutting out the old switch and connecting a new switch through a conduit and junction box. However, the area had not been adequately surveyed prior to the job commencing, when the assigned RP technician failed to survey and inform the workers of the sump room floor contamination levels. The associated survey levels (i.e., max beta-gamma smears of 28 millirad/hr/100cm2, which would indicate greater than 100,000 dpm/100cm2) identified supported the Rule of Three, but adequate RP controls were not implemented. This resulted in at least two uptakes exceeding 10 millirem CEDE. It was determined this issue was primarily attributed to a failure to conduct adequate surveys in a work area. This was documented as a violation of NRC and licensee requirements in which additional details are provided.

AR 470470: This contamination event occurred on May 12, 2025, when two workers received uptakes while performing work in the heater bay area to remove an actuator from a valve (MD-MO-71). When NRC discussed the issue with workers, it was noted that there were prior steam leaks into the valve associated piping insulation that increased the contamination levels in the work area. The pre-job surveys, dated May 4, 2025, used to brief the workers performing the tasks were not adequate for the work conditions. The workers needed to crawl through tight spaces to get to the work location and ultimately were exposed to conditions that were not adequately surveyed. Post survey results, dated May 12, 2025, showed the contamination levels were considerably higher than expected. This resulted in at least two uptakes for the valve technicians, but no unplanned internal doses exceeded 10 millirem CEDE.

AR 471180: This contamination event occurred on May 24, 2025, when a worker received an uptake while performing work on the 441-foot level of the turbine building. Specifically, laborers were cleaning the condenser false bottom bay B while wearing double sets of protective clothing, a hood, and a face shield, as required by the assigned RWP. However, the communication on what debris was required to be retrieved was lacking. There was no requirement for continuous coverage based on the known radiological conditions, so the licensee suspected that finer debris was collected, than what was briefed, and this finer debris somehow entered beneath the face shield resulting in facial contamination. This resulted in at least one workers uptake, but no unplanned doses exceeded 10 millirem CEDE. Per NRCs review, the causes related to this event were not deemed related to the May 2021 CAPRs.

Collectively reviewing these events, which resulted in numerous worker uptakes of radioactive contaminants due to licensee failures to implement adequate RP controls and adequate risk mitigation, at least three of them involved a performance deficiency associated with inadequate CAPRs implemented from the May 2021 contamination uptake event.

Although the fourth uptake event did not relate to the CAPRs, it still highlights the need for enhanced RP fundamentals training with workers/laborers, and the importance of three-way communication.

As a result of the May 2021 contamination uptake event, two updated CAPRs related to root causes were implemented. This involved CAPR 1.1 - dynamic learning activities to enforce RP standards and requirements, positive RP command of the work, and control of radiological work activities; and

(2) CAPR 3.1-updating procedures to enhance and define risk mitigation and elimination actions for work activities. When the NRC inspector reviewed the licensees RCE, it stated that corrections were provided to procedures PPM 11.2.2.12, 11.2.2.14, and Form 26840 for risk work plans. PPM 11.2.2.12, Radiological Risk Assessment and Management, was updated to instruct the job planner to not only define risk mitigation and elimination, but to determine the initial risk assuming no mitigation or elimination actions, and to only allow the risk categorization to credit an action that eliminates the risk. These RP enhancements would require more rigor in the licensees job planning. In addition, contamination levels of greater than 100,000 dpm/100cm2 would require the licensees Rule of Three controls, which include

(1) one form of engineering controls, (2)respiratory protection or a face shield, and

(3) air sampling.

In the RCE for the May 2021 contamination uptake event, the licensee noted that the decision to classify the job as medium risk underestimated the potential dose risk and resulted in a reduced level of radiological controls and barriers. Also, the decision to not use the Rule of Three compromised defense in depth and was a missed opportunity to have more robust controls in place to minimize radiological risk.

During the May 2025 uptake events, in the uptake event related to AR 470380470380 with unplanned uptakes exceeding 10 millirem CEDE, the assigned risk to the job activity was assigned as medium risk although breaching of the system and the associated survey data without the assumption of mitigation or elimination would have supported a high risk RWP.

Specifically, the licensee noted that they did not implement more rigorous controls with the maintenance of highly contaminated MSRVs because the workers were not expected to be at the plane of the valve breach, eliminating the radiological risk. The PPM 11.2.2.12 updates state the job planning is to determine the initial risk assuming no mitigation or elimination actions.

Also, the Rule of Three was not adequately implemented based on the follow-up survey levels associated with AR 470380470380and AR 470395470395(beta-gamma contamination of greater than 100,000 dpm/100cm2). It is understood that the associated contamination levels may not have been known at the time of pre-job planning due to the course of the job activity, but even as additional areas where work commenced were breached or accessible, work was not stopped to appropriately assess/survey the actual radiological conditions and then re-evaluated to plan the remainder of the job safely, including updating the associated radiological risk for the job activity and/or ensuring the appropriate RP controls were in place.

In essence, more rigorous job planning was lacking.

In addition to the information provided that demonstrated the failure to survey the work area was inadequate in some instances, NRC also determined that the RWP requirements were not followed as required for some job activities. Specifically, the requirement(s) to wear appropriate protective equipment (i.e., face shields and/or respiratory protection) were not adequately implemented. This is even after the workers inquired about the use of wearing such equipment based on prior work history. These RWP requirements also included continuous RP coverage for at least one of the work activities, and NRC determined this was not adequately implemented. These performance deficiencies are all reflective of the causes related to the White performance issues during the May 2021 contamination uptake event, which resulted in related CAPRs.

Thus, by NRCs review, it was concluded that the licensees May 2021 uptake event CAPRs to enforce RP requirements, enhance positive RP command and control of work activities, and enhance risk assigned to work activities were not effectively implemented during the aforementioned job activities during the May 2025 outage.

Corrective Actions: As immediate corrective actions, the licensee entered this issue into their corrective action program and implemented revisions of the root cause evaluation which added corrective actions to preclude repetition for contamination uptake event performance issues, a significant condition adverse to quality.

Corrective Action References: AR 00478737

Performance Assessment:

Performance Deficiency: The licensee failed to establish measures to assure significant conditions adverse to quality, such as failures to perform adequate surveys, failures to follow the RWP requirements, failures to use appropriate respiratory protection, and failures to assign the appropriate risk to work, as identified for a White finding, were promptly and adequately corrected, such that actions were taken to preclude repetition.

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. Additionally, there was a potential this performance deficiency could lead to a more significant radiation safety concern because of an ineffective radiation program barrier. Specifically, a failure to implement prompt and effect corrective actions for prior contamination uptake events resulted in a series of similar events causing unplanned internal doses to workers.

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

(1) it was not a finding in ALARA Plans or work controls,

(2) it was not an overexposure,

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

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

Cross-Cutting Aspect: P.5 - Operating Experience: The organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee did not utilize internal operating experience from their May 2021 contamination uptake event to adequately implement RP controls, perform adequate surveys, use appropriate respiratory protection or a face shield, and assign the correct risk to job activities during their R27 outage. This resulted in multiple contamination uptake events, within a period of 15 days, and unplanned internal dose to numerous workers.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall ensure that the cause of the condition is determined, and corrective action taken to preclude repetition.

Contrary to the above, between May 9 - 24, 2025, the licensee did not assure that measures established to ensure that the causes of the significant conditions adverse to quality, such as deficiencies and nonconformances, were corrected to prevent recurrence. Specifically, corrective actions focused on the deficiencies developed from a significant condition adverse to quality White performance issue were not implemented effectively to preclude repetition.

Consequently, the licensee repeated performance deficiencies involving inadequate surveys, failures to follow RWP requirements, failures to use appropriate respiratory protection, and failures to assign the appropriate risk to work activities. These performance deficiencies resulted in multiple contamination uptake events and unplanned internal doses to workers.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

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

• On December 11, 2025, the inspectors presented the integrated inspection results to Jeremy Hauger, acting Site Vice President and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.04

Drawings

M524-1

Standby Service Water System

140

71111.04

Drawings

M524-2

Standby Service Water System

24

71111.04

Procedures

OSP-SW-M103

HPCS Service Water Valve Position Verification

71111.04

Procedures

SOP-RHR-STBY

Placing RHR in Standby Status

71111.05

Fire Plans

CGS PRE-FIRE PLAN

Turbine Generator 501'

71111.05

Fire Plans

Fire Area TG1/2, Zone

TG-12

Turbine Generator 441' RB-RW-DG Corridor

71111.05

Fire Plans

PFP-RW-467

RADWASTE 467

71111.12

Procedures

ICP-RCIC-Q901

RCIC Isolation on RCIC Steam Supply Flow High DIV 2-

CFT/CC

71111.12

Work Orders

WO 2108685, 2163722, 2178933, 2207751

71111.13

Drawings

FM892-1

Fire Area Boundary Plan - Ground Floor

71111.13

Miscellaneous

FPEP-25-1002

Fire Protection Evaluation Permit

71111.13

Miscellaneous

PSA-FIRE-A4-0001

Maintenance Rule Fire A4 Risk Assessment

71111.13

Procedures

1.3.83

Protected Equipment Program

71111.13

Procedures

1.3.85

On-Line Fire Risk Management

71111.15

Corrective Action

Documents

474127

71111.15

Corrective Action

Documents

Resulting from

Inspection

00473171, 00473187, 00473189, 00473416, 00473667,

00473714, 00473716, 00473717, 00473719, 00473722,

00473802, 00473821, 00473858, 00473929, 00473948,

00473989, 00474045, 00474049, 00474167, 00474230,

00474376, 00474470, 00474491, 00474497, 00474514,

00474580, 00474615, 00474634, 00474643, 00474707,

00474746, 00474756, 00474757, 00474758, 00474760,

00474763, 00474764, 00474765, 00474768, 00474793,

00474799, 00474802, 00474803, 00474847, 00474850,

00474851, 00474886, 00474899, 00474902, 00474916,

00474960, 00475019, 00475031, 00475047, 00475048,

00475050, 00475082, 00475092, 00475168, 00475191,

00475201, 00475209, 00475264, 00475270, 00475285,

00475287, 00475343, 00475480, 00475483, 00475540,

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

00475692, 00475723, 00475843, 00475890, 00475902,

00475969, 00476056, 00476071, 00476086

71111.15

Procedures

10.25.181

Single Cell Charging of Batteries

71111.15

Work Orders

WO 211388

71111.24

Corrective Action

Documents

474899

71111.24

Engineering

Changes

19595

Replacement of the RRC Adjustable Speed Drives

71111.24

Engineering

Changes

EC0000019595R001S

Field Change Request

1S

71111.24

Procedures

10.2.101

Informational Visual Examinations

71111.24

Procedures

ISP-RPS-B603

Reactor Vessel Steam Dome Pressure High-RPS Trip

Channel A2 RTT

71111.24

Procedures

OSP-CCH/IST-M702

Control Room Emergency Chiller System B Operability

71111.24

Procedures

TSP-RCIC-B801

RCIC Leakage Surveillance

71111.24

Work Orders

WO 2069781, 2206650, 2207838, 2210427

71152A

ALARA Plans

RWP 30005133

R27 MSRV Replacement and Associated Work ALARA

Plan

71152A

Corrective Action

Documents

468837, 468887, 469198, 469263, 470380, 470395,

470470, 471180

71152A

Miscellaneous

HMIS Intake and Internal Dose Evaluation Report for

Energy Northwest Nuclear Power Plant

07/22/2025

71152A

Miscellaneous

RSCS Energy Northwest Independent Internal Dose

Calculation for the May, 2025 Radioactive Material Intake

Events

08/27/2025

71152A

Miscellaneous

Timeline of Events for May 2025 Uptakes and Bioassays

08/18/2025

71152A

Miscellaneous

TEDE-ALARA Evaluation for RWP 30005133

05/10/2025

71152A

Miscellaneous

Energy Northwest Prompt Investigation Report for AR-

CR 00470380

06/11/2025

71152A

Miscellaneous

Energy Northwest Prompt Investigation Report for AR-

CR 00470395

05/15/2025

71152A

Procedures

PPM 1.3.76

Integrated Risk Management

71, 72

71152A

Procedures

PPM 11.2.13.1

Radiation and Contamination Surveys

71152A

Procedures

PPM 11.2.4.6

Form 27297: Energy Northwest Bioassay Worksheet

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71152A

Radiation

Surveys

M-20250509-12

R27 DW 548' MSRV 4B / Follow up

05/09/2025

71152A

Radiation

Surveys

M-20250510-18

R27 EDR Sump Floor / Follow up

05/10/2025

71152A

Radiation

Surveys

M-20250510-6

R27 DW 548 MSRV 4A Breach and Removal

05/10/2025

71152A

Radiation Work

Permits (RWPs)

30005133

R27 Drywell (DW) Main Steam Relief Valve (MSRV)

Maintenance LHRA

71152A

Radiation Work

Permits (RWPs)

30005194

R27 Radiologically Controlled Area (RCA) HRA

1