IR 05000397/2025002
| ML25213A155 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/11/2025 |
| From: | Patricia Vossmar NRC/RGN-IV/DORS/PBA |
| To: | Schuetz R Energy Northwest |
| Bywater R | |
| References | |
| IR 2025002 | |
| Download: ML25213A155 (1) | |
Text
August 11, 2025
SUBJECT:
COLUMBIA GENERATING STATION - INTEGRATED INSPECTION REPORT 05000397/2025002
Dear Robert Schuetz:
On June 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Columbia Generating Station. On July 29, 2025, the NRC inspectors discussed the results of this inspection with Reginald Wainwright, Plant General Manager, and other members of your staff. The results of this inspection are documented in the enclosed report.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. 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, Patricia J. Vossmar, Chief Reactor Projects Branch A Division of Operating Reactor Safety Docket No. 05000397 License No. NPF-21
Enclosure:
As stated
Inspection Report
Docket Number:
05000397
License Number:
Report Number:
Enterprise Identifier:
I-2025-002-0010
Licensee:
Energy Northwest
Facility:
Columbia Generating Station
Location:
Richland, WA
Inspection Dates:
April 1, 2025, to June 30, 2025
Inspectors:
J. Brodlowicz, Resident Inspector
R. Bywater, Senior Project Engineer
N. Greene, Senior Health Physicist
C. Highley, Senior Resident Inspector
J. Mejia, Reactor Inspector
B. Tharakan, Technical Assistant
Approved By:
Patricia J. Vossmar, Chief
Reactor Projects Branch A
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
Reactor Coolant System Pressure Boundary Leakage through a Valve Bonnet Vent Line Weld Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000397/2025002-01 Open/Closed Not Applicable 71111.15 A self-revealed, Severity Level IV non-cited violation of Technical Specification 3.4.5, RCS Operational Leakage, was identified when the licensee operated Columbia Generating Station with reactor coolant pressure boundary leakage through a weld on the bonnet vent line of reactor recirculation (RRC) valve RRC-V-67B, from July 1, 2023, to April 12, 2025.
Non-Functional Installed Continuous Particulate Monitoring System Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025002-02 Open/Closed
[H.6] - Design Margins 71124.03 The inspectors identified a Green non-cited violation of 10 CFR 50.34 for the failure to maintain a functional installed continuous particulate monitoring system in accordance with the Final Safety Analysis Report. Specifically, 10 CFR 50.34(b) requires that a Final Safety Analysis Report shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole. Additionally, 10 CFR 50.34(f)(2)(xxvii) requires the licensee to provide for monitoring of in-plant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions. Chapter 12.3.4.2.b of the Final Safety Analysis Report describes the airborne radiation monitors, including the installed continuous particulate monitoring system, which was determined to be non-functional for a significant period of time between August 15, 2017, and April 23, 2025.
Additional Tracking Items
None.
PLANT STATUS
The unit began the inspection period at 87 percent rated thermal power and coasting down in support of the scheduled refueling outage. On April 11, 2025, the unit shut down for the R27 refueling outage. On June 14, 2025, the unit started up from the R27 refueling outage. On June 19, 2025, the unit reached 85 percent rated thermal power. On June 20, 2025, the unit was down powered to 43 percent due to feed water heat exchanger and moisture separator concerns. On June 21, 2025, the unit restored rated thermal power to 80 percent. On June 23, 2025, the unit was shut down to perform a balance shot on the main turbine. On June 28, 2025, the unit was restarted. On June 30, 2025, the unit was returned to rated thermal power 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 on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
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)stator water cooling shutdown lineup for turbine outage work on April 28, 2025 (2)standby gas treatment train A on May 18, 2025 (3)reactor core isolation cooling partial lineup prior to startup on June 11, 2025
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
(1)high-pressure core spray system on May 22, 2025
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) R-310 reactor building reactor steam tunnel on April 21, 2025 (2)fire area R-2/U, reactor 501-foot elevation, containment on April 24, 2025 (3)turbine generator 441 spaces on April 28, 2025 (4)fire area RC-10/U, main control room, on May 13, 2025 (5)fire area R-1, reactor 606-foot elevation, refueling floor, on May 28, 2025 (6)north and south heater bay fire area TG-1/2 on June 2, 2025
- (7) TG-1/2, 501-foot elevation, low pressure feed water heater bay and main generator area on June 5, 2025
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01)
The inspectors evaluated boiling-water reactor nondestructive testing by reviewing the following examinations from April 21 - May 21, 2025:
(1)
1. Nondestructive Examinations
a.
Ultrasonic Examinations i.
main steam, 26MS(1)C-8, pipe to elbow ii.
main steam, 26MS(1)C-10, pipe to elbow iii.
main steam, 26MS(1)C-16, pipe to valve b.
Dye Penetrant Examinations i.
main steam, 2MS(9)D-2, pipe to elbow ii.
main steam, 2MS(9)D-3, elbow to pipe c.
Visual Examinations i.
main steam, MS-HA-2, spring support ii.
metal containment, Shell-501-0-G, shell 0-90 iii.
metal containment, Shell-501-90-G, shell 90-180 iv.
metal containment, Shell-501-180-G, shell 180-270 v.
metal containment, Shell-501-270-G, shell 270-360 d.
Magnetic Particle Examinations i.
main steam, MS-HA-2(W), welded lugs
2. There were no relevant indications accepted for continued service
3. Welding Activities
a.
Work Order 2167077, seal weld, RHR-V-84A
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 a reactor shutdown and cooldown for a refueling outage on April 13, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated just in time training (JITT) for reactor shutdown on April 2, 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)shutdown risk plan for the R-27 refueling outage on April 9, 2025
- (2) Yellow shutdown risk calculation and protected equipment with division 1 secured during week of April 20th, on April 30, 2025
- (3) Yellow shutdown risk calculation and protected equipment with partial division 1 and 2 secured during week of May 12th, on May 16, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)main steam relief valve failures of MS-RV-2D, 5C, 5B, 1D, Condition Report 468564,
===468565, 468563, and 468562, on April 12, 2025 (2)loss of B residual heat removal (RHR) in shutdown cooling mode, Condition Report 468574, on April 12, 2025 (3)failure of cleanliness 'C' controls, diesel generator fuel oil tank DG-FO-Tk, 1A, Condition Report 469840, on May 1, 2025
- (4) RCIC-V-90 operability following OSP-RCIC/IST-Q701 surveillance testing on May 19, 2025
- (5) OSP-RHR/IST-R704 RHR valve operability on May 21, 2025 (6)inboard main steam isolation valve (MSIV) 28D exceeding technical specification allowed limits for leakage completed on June 17, 2025 (7)missed technical specification surveillance for reactor recirculation pump trip - turbine generator valve fast closure, ISP-RPS-S902, RPS-PS-5C, Condition Report 472918, on June 25, 2025
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
=
The inspectors evaluated the following temporary or permanent modifications:
- (1) RCIC-P-3 replacement completed on June 11, 2025 (2)temporary charger E-C1-7B installation completed on June 11, 2025
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated R27 refueling outage activities from April 10, 2025, to June 29, 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) (5 Samples)
(1)shutdown cooling valve operability following repairs on May 21, 2025 (2)as left leak rate test for MS-V-28D MSIV on May 30, 2025
- (3) Work Order 2215507 leak check post maintenance test following repair of RHR-V-50A on June 3, 2025
- (4) Work Order 2223416 reactor core isolation cooling post maintenance testing prior to 150 psig, on June 15, 2025 (5)as left leak rate test TSP-MSIV-B801 for MSIV 28D on June 17, 2025
Surveillance Testing (IP Section 03.01) (4 Samples)
- (1) TSP-LPCS/RHRA-B502 LPCS/RHR A annunciator surveillance on May 19, 2025
- (2) OSP-INST-B702 alternate remote shutdown panel operability surveillance on May 30, 2025 (3)dry well/wet well bypass leak rate test, TSP-CONT-B801, on June 9, 2025 (4)main turbine bypass valve testing, OSP-MS-Q702, on June 17, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)main steam relief valve testing, ISP-MS/IST-R101, Work Order 2207138, on April 12, 2025
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) TSP-RRC/X77AA-C801 leak rate testing of RRC-V 19 and 20 on May 30, 2025
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors evaluated the following submitted Emergency Action Level and Emergency Plan changes.
1. Emergency Plan, Columbia Generating Station, Revision 69 (NRC notified on
April 9, 2025)
2. Plant Procedures Manual 13.4.1, "Emergency Notifications," Revision 46
(NRC notified on April 24, 2025)
3. Plant Procedures Manual 13.1.1A, "Classifying the Emergency - Technical
Bases," Revision 35 (effective date March 27, 2025; NRC notified on May 14, 2025)
Several screenings and evaluations associated with the 10 CFR 50.54(q) emergency plan change process were reviewed as well. These evaluations do not constitute NRC approval.
RADIATION SAFETY
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, identifies the concentrations and quantities of radioactive materials, and assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
(1)the licensee's process of conducting radiological surveys of potentially contaminated items leaving the radiologically controlled area (RCA)
(2)the licensee's process of surveying workers exiting the RCA during the licensee's R27 refueling outage (3)radiological controls for radioactive material and radwaste trash stored in the radwaste building during the licensee's R27 refueling outage
Radiological Hazards Control and Work Coverage (IP Section 03.04) (5 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Radiation Work Permit (RWP) 30005194, R27 Operations in High Radiation Area (HRA)/ High Contamination Area (HCA), April 21, 2024
- (2) RWP 30005117, R27 Undervessel (U/V) Support Work, April 22-23, 2025
- (3) RWP 30005118, R27 Control Rod Drive Mechanisms (CRDM) U/V Remove and Replace 20 CRDMs, April 22-24, 2025
- (4) RWP 30005137, R27 Drywell / Steam Tunnel Lock High Radiation Area (LHRA)
ISI/Scaffold/Insulation, April 23, 2025
- (5) RWP 30005140, R27 CRDM Positioner Indicator Probe (PIP) Replacements, April 24, 2025 High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
(1)437-foot elevation of radwaste building, resin liner NUPAC cage (2)437-foot elevation of radwaste building, north, center, and south shield door (3)522-foot elevation of reactor building, north pipe space (4)522-foot elevation of reactor building, reactor water clean-up rooms A and B (5)548-foot elevation of reactor building, reactor water clean-up heat exchanger room Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Permanent Ventilation Systems (IP Section 03.01) (2 Samples)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
(1)reactor and radwaste building ventilation systems (2)control room emergency filtration system
Temporary Ventilation Systems (IP Section 03.02) (2 Samples)
The inspectors evaluated the configuration of the following temporary ventilation systems:
(1)606-foot elevation of the reactor building, refuel continuous air monitor high efficiency particulate air filtration system (2)residual heat removal valve repair high efficiency particulate air filtration system
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===
- (1) April 1, 2024, through March 31, 2025
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) July 1, 2024, through March 31, 2025 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, 2024, through March 31, 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 identified that a fire protection design feature of the emergency diesel generator day tank rooms was inconsistent with its description in the operating license safety evaluation report. The issue is documented in this report as a Very Low Safety Significance Issue Resolution Item.
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 the transient combustibles program that might be indicative of a more significant safety issue. From August 2024 through February 2025, there were four NRC identified condition reports regarding transient combustible permits (TCPs) with only one licensee identified condition report in the same time frame. Since March 2025, Columbia performed a sweep, generated at least seven condition reports and identified at least six instances of the TCP procedure not being followed with the NRC identifying two additional cases. The trend appears to be improving, still with room for improvement, and is not indicative of a more significant safety issue at this time.
- (2) The inspectors reviewed the licensees corrective action program to identify potential trends in Quality Control/Quality Assurance (QC/QA) program that might be indicative of a more significant safety issue. The inspectors identified an adverse trend in the QC/QA program based on condition reports. The observation was shared with the licensee.
Condition Reports: 468531, 468802, 468846, 468859, 468985, 469295, 469389, 470067, 470135, 470214, 470343, 470350, 470457, 470845, 470858, 470888, 471060, 471845, 471897, 471928,
INSPECTION RESULTS
Reactor Coolant System Pressure Boundary Leakage through a Valve Bonnet Vent Line Weld Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000397/2025002-01 Open/Closed Not Applicable 71111.15 A self-revealed, Severity Level IV non-cited violation of Technical Specification 3.4.5, RCS Operational Leakage, was identified when the licensee operated Columbia Generating Station with reactor coolant pressure boundary leakage through a weld on the bonnet vent line of reactor recirculation (RRC) valve RRC-V-67B, from July 1, 2023, to April 12, 2025.
Description:
On July 1, 2023, Columbia received indications of a reactor coolant system (RCS) leak when both containment monitoring system (CMS) radiation ratemeters were reading higher than normal and similarly tracking. However, other parameters such as equipment and floor drain leakage, containment pressure and temperature, and isotopic analysis of containment atmosphere were steady and did not indicate symptoms of an RCS leak. As samples were collected and analyzed over the following weeks, other indications revealed that a small RCS leak was present, such as:
1. analysis of filters in the containment radioactive particulate and radiation monitoring
system indicating short-lived radioisotopes
2. CMS-RIS-12A/2 and 12B/2 spiking high several times per hour
3. analysis of floor drain radioactive (FDR) sump samples indicating short-lived
radioisotopes
4. anomalous indications of increased leakage detected by drywell floor drain
instrumentation On July 17, 2023, in response to these symptoms, the licensee initiated an operational decision-making issue (ODMI) based on AR/CR 448527, which was written to address the need to analyze trends indicating a potential RCS leak in containment. The ODMI established monitoring parameters for RCS leakage via station Procedure SOP-FDR-OPS, Manual Determination of Leakage, and isotopic analysis of FDR sampling and CMS-SR-20, -21 particulate filters. The ODMI also developed action triggers for increases in RCS unidentified leakage. An accompanying decision-making matrix was used to assist in determining an appropriate path forward to address an increasing RCS unidentified leakage trend. Initially, a benefit risk evaluation of the identified options concluded the best option to pursue was Continue monitoring leakage per approved ODMI.
On October 7, 2024, the decision-making matrix was revisited to reconsider performing a down power with containment entry to identify the leak location. The station decided again to Continue monitoring leakage per approved ODMI, but expanded the option to include Perform low power drywell entry and inspection prior to R27. The licensee continued to monitor RCS leakage throughout the operating cycle and at no time did the leakage rates challenge technical specification operational leakage rate limits for identified or unidentified leakage.
On April 11, 2025, just prior to entering R27 refueling outage and at 15 percent power, station personnel conducted a low-power drywell entry and inspection. A steam plume was identified coming from a 3/4 vent line on RRC-V-67B, which correlated to a 0.05 gpm leak rate. After entering MODE 3, Operations declared this to be RCS pressure boundary leakage and entered technical specifications (TS) Limiting Condition for Operation (LCO) 3.4.5, Action C (enter MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />). MODE 4 was entered on April 12, 2025.
Corrective Actions: Columbia performed a root cause analysis and determined the bonnet of the valve was not designed to prevent high cyclic fatigue of the bonnet vent weld. During the spring 2025 refueling outage, the station repaired the RRC-V-67B bonnet vent weld leak and also replaced other welds identified as at-risk under WO 2228369-08. The licensee also performed an extent of condition review on similar small-bore piping on lines identified as being susceptible to cyclic fatigue. A total of 56 locations on reactor recirculation system A and B were inspected. A redesign of the RRC-V-67B vent line to sufficiently address the cyclic fatigue issue (CAPR 1.1 and 1.2) and a plant modification for the other 56 locations that identified a fatigue issue (EOC2.1, 2.2, and 2.3), was scheduled for implementation during the next refueling outage in June 2027.
Corrective Action References: CR/AR 00468526
Performance Assessment:
The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.
Specifically, the indication of the leak by the containment radiation monitors, the magnitudes of the unidentified and identified leakage rates, and the radiochemical analysis results remaining steady after the initial indication in July 2023 were consistent with previous station and industry operating experience of being a leak from a non-pressure boundary location.
Enforcement:
The ROPs significance determination process does not specifically consider violations without associated performance deficiencies in its assessment of licensee performance. Therefore, it is necessary to address this violation without a performance deficiency using traditional enforcement to adequately deter non-compliance.
Severity: The NRC Enforcement Policy, Section 2.2.1, states, in part, whenever possible, the NRC uses risk information in assessing the safety significance of violations. Accordingly, after considering that the condition represented very low safety significance, the inspectors concluded that the violation would be best characterized as Severity Level IV under the traditional enforcement process. For information, the inspectors screened the significance of the condition using IMC 0609 Appendix A, "The Significance Determination Process for Findings At-Power," and determined that the condition was of very low safety significance (Green) because the very small leak did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident and did not affect systems used to mitigate a loss of coolant accident.
Violation: Columbia Generating Station, TS Limiting Condition for Operation (LCO) 3.4.5 requires that while the plant is operating in MODES 1, 2, and 3, RCS operational leakage shall be limited to no pressure boundary leakage. Technical Specification Action Statement 3.4.5.C requires that if pressure boundary leakage exists, the unit shall be placed in MODE 3 and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively.
Contrary to the above, from July 1, 2023, to April 12, 2025, while the plant was operating in MODE 1, pressure boundary leakage existed, and the unit was not placed in MODE 3 and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively. Specifically, indications of RCS leakage were present on July 1, 2023, and the unit remained operating in MODE 1 until a plant shutdown on April 11, 2025, when the licensee discovered the indications were the result of pressure boundary leakage through a weld on the bonnet vent line of reactor recirculation valve RRC-V-67B.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Non-Functional Installed Continuous Particulate Monitoring System Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000397/2025002-02 Open/Closed
[H.6] - Design Margins 71124.03 The inspectors identified a Green non-cited violation of 10 CFR 50.34 for the failure to maintain a functional installed continuous particulate monitoring system in accordance with the Final Safety Analysis Report. Specifically, 10 CFR 50.34(b) requires that a Final Safety Analysis Report shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole. Additionally, 10 CFR 50.34(f)(2)(xxvii) requires the licensee to provide for monitoring of in-plant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions.
Chapter 12.3.4.2.b of the Final Safety Analysis Report describes the airborne radiation monitors, including the installed continuous particulate monitoring system, which was determined to be non-functional for a significant period of time between August 15, 2017, and April 23, 2025.
Description:
As described in Chapter 12.3.4.2.b and 12.3.4.4 of the licensee's Final Safety Analysis Report (FSAR), Columbia Generating Station has a permanently installed continuous particulate monitoring system (ICPMS) that was designed for responsive personnel protection and plant surveillance. There are three airborne radiation monitors within the ICPMS that measure the airborne particulate activity levels in the radwaste and reactor building ventilation exhaust and furnish recording signals to the main control room.
One airborne radiation monitor (RRA-RIS-3) is located in the reactor building and the other two (WRA-RIS-1 and WRA-RIS-2) are located in the radwaste building.
In the licensee's FSAR, corrective action program, and other documentation (e.g.,
procedures) various terms are used to describe the airborne radiation monitors in the ICPMS.
For purposes of consistency in this report, the three airborne radiation monitors will be collectively referred to as the ICPMS.
As designed, each of the three ICPMS units draw approximately three
- (3) cubic feet per minute air samples from their respective ventilation system through a particulate filter which is monitored by a shielded beta detector with an efficiency of approximately 30 percent. The resultant response of the system is an increase of about 350 counts per minute for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of sampling at a 1.0E-10 microcuries/cm3 concentration. External gamma radiation will increase the background by 70 counts per minute/millirem per hour.
The FSAR states that the ICPMS is reasoned to be the most responsive personnel protection and internal plant surveillance mechanism available. In addition, all tasks with potential for generating airborne contamination will be performed only when authorized by a radiation work permit (RWP). The RWP documents the assessed radiological hazards, establishes additional monitoring and sampling requirements, and if necessary, specifies required engineering control and/or respiratory protection. During outages, the above airborne monitoring system will be augmented by additional iodine sampling (continuous and grab) on the refueling floor since airborne iodine concentrations are known to become significant at this time.
On April 23, 2025, during refueling outage 27, the inspectors walked down the ICPMS and observed that none of the three airborne particulate monitors were in operation. The inspectors observed that one monitor in particular, RRA-RIS-3, had a deficiency tag on it from 2017. The inspectors further inquired about the status of repairs to RRA-RIS-3, and the other monitors. The inspectors also reviewed several condition reports that documented various issues with the ICPMS, including failed operational checks, high and low flow rate alarms, and the inability to calibrate the monitors. Since the inspectors observed a deficiency tag from 2017, they requested the licensee to provide a timeline of ICPMS outages and repairs since 2017. The licensee provided a timeline to the inspectors indicating RRA-RIS-3 had been out of service since August 15, 2017, which corresponded to the deficiency tag on the monitor observed by the inspectors in the field.
Additionally, the two ICPMS monitors in the radwaste building were also out of service for various lengths of time since January 1, 2017. WRA-RIS-2 was only in-service for about 3 months since 2017 and had been out of service since August 4, 2021. WRA-RIS-1 was in-service for almost the entire time since January 1, 2017, except for about 2 months (1 month in 2017 and 1 month in 2021). Although at the time of the walkdown on April 23, 2025, the inspectors found WRA-RIS-1 without power due to operations personnel de-energizing the electric panel supplying power to the monitor for planned electrical maintenance during the refueling outage. Radiation protection supervision was not informed about the loss of power to WRA-RIS-1, nor was a condition report initiated, or a health physics logbook entry made for the loss of power per plant procedures. The licensee generated Action Request 469319 to address this issue.
The inspectors evaluated the information provided by the licensee and determined that the ICPMS requirements described in the FSAR had not been met for several months to several years, in the case of RRA-RIS-3. Although the licensee had several portable continuous air monitors permanently stationed around the plant, the loss of ICPMS functions had not been compensated or adequately evaluated by the licensee to determine if the sensitivity, efficiency, and accuracy provided by ICPMS could be replaced such that the effectiveness of airborne radioactivity monitoring in the plant had not been jeopardized. The inspectors determined this was a violation of the licensee's safety analysis and processed the finding in accordance with the NRC's reactor oversight program.
Corrective Actions: The licensee will perform an evaluation to determine the number of portable continuous air monitors that will be required to adequately compensate for the installed continuous particulate airborne radiation monitors to meet the safety analysis specifications and will address the repair or replacement of the monitors through their commercial change package review process.
Corrective Action References: AR 469351469351 449789, 469319
Performance Assessment:
Performance Deficiency: The failure to maintain a functional ICPMS in accordance with the FSAR is a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. A nonfunctioning ICPMS could allow fractional increases in airborne radioactivity concentrations in the reactor and radwaste buildings to go undetected for longer periods of time without the appropriate countermeasures deployed or evaluated to prevent unintended occupational doses.
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.6 - Design Margins: The organization operates and maintains equipment within design margins. Margins are carefully guarded and changed only through a systematic and rigorous process. Special attention is placed on maintaining fission product barriers, defense-in-depth, and safety-related equipment. Specifically, from August 15, 2017, to April 23, 2025, the ICPMS was not functioning within the FSAR design margins, and the licensee failed to perform a safety evaluation or identify compensatory measures to substitute for the accuracy, efficiency, and sensitivity of the ICPMS.
Enforcement:
Violation: Title 10 CFR 50.34(b) requires, in part, a Final Safety Analysis Report shall include information that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole. Additionally, 10 CFR 50.34(f)(2)(xxvii) requires the licensee to provide for monitoring of in-plant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions.
Contrary to the above, between August 15, 2017, and April 23, 2025, the licensee failed to present a safety analysis on the operating status of the ICPMS and provide for monitoring of in-plant radiation and airborne radioactivity as appropriate for routine work and accident conditions in the reactor and radwaste buildings. Specifically, the licensee operated the plant with a non-functional ICPMS with no compensatory actions or safety evaluation to demonstrate airborne radioactivity is measured with the same accuracy, efficiency, and sensitivity as the ICPMS, as described in the FSAR.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Very Low Safety Significance Issue Resolution Process: Question on Diesel Generator Day Tank Room Fire Protection Configuration 71152A This issue is a current licensing basis question and inspection effort is being discontinued in accordance with the Very Low Safety Significance Issue Resolution (VLSSIR) process. No further evaluation is required.
Description:
The inspectors identified an issue of concern that the as-built configuration of the emergency diesel generator day tank rooms was inconsistent with information contained in the NRCs operating license safety evaluation report for the licensees fire protection program.
A portion of the NRCs safety evaluation of the licensees fire protection program is documented in NUREG-0892, Safety Evaluation Report related to the operation of WPPNS Nuclear Project No. 2, March 1982. This document included the following relevant information:
Each diesel generator and day tank is protected by a pre-action sprinkler system.
Day tanks, each having a 3000-gal capacity, are provided in a separate enclosed area. One tank is provided for each diesel generator. The day tank enclosures have a minimum fire resistance, including doors, of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Enclosure penetrations are sealed. The day tank areas are vented to avoid the accumulation of oil fumes. The enclosures are capable of containing the entire contents of the day tanks and suppressant water. No floor drains are provided in the day tank rooms.
Our guidelines of BTP [Branch Technical Position] CMEB [Chemical Engineering Branch] 9.5-1 recommend that a maximum of 1100 gal be permitted for diesel generator day tanks. The total volume of 3000 gal in addition to any feasible amount of fire suppressant water will be contained within the day tank room. The staff finds that the day tanks and their enclosed areas are an acceptable deviation to from (sic) BTP 9.5-1, Section C.7.i.
During a fire protection system inspection of the day tank rooms (Fire Areas DG-7, DG-8, and DG-9), the inspectors noted that each day tank room was accessed by climbing steps and entering through a 3-hour rated fire door that was approximately 30 inches above the floor of the diesel generator building. Additionally, the inspectors noted that each day tank room had a 3-hour rated fire damper in the wall between the day tank room and diesel generator room, approximately 30 inches above the diesel generator building floor, held open by a fusible link. Neither the fire door, nor the fire damper, were flood barriers. Because the day tank rooms had a high-density pre-action fire suppression system and contained no floor drains, the inspectors identified an issue of concern whether the enclosure was capable of containing the day tank fuel and fire suppressant water without overflowing from the day tank room and flooding the diesel generator room. This condition appeared contrary to the configuration approved by the NRC staff in the safety evaluation report referenced above.
The inspectors reviewed the pre-fire plans for the diesel generator day tank rooms and noted that each plan stated, A major Diesel Fuel leak accompanied by a fire and A/S [automatic suppression] system actuation could present the potential for an oil overflow from the room within one minute. The inspectors also reviewed Calculation ME-02-03-02, Calculation for Diesel Generator Building Flooding Analysis, Revision 1, which determined flooding heights in diesel generator building rooms in the event of a moderate energy line crack, and Calculation ME-02-02-40, PFSS Flooding Analysis - Diesel Generator Building, Revision 0, which determined flooding heights in diesel generator building rooms in the event of a fire with automatic and manual fire suppression. Each calculation accounted for water flowing through the door opening, but neither calculation identified the fire damper as a flow path from the day tank enclosure into the diesel generator room.
After discussing the identified concern with the inspectors, the licensee entered the issue into the corrective action program. The inspectors did not have an immediate safety concern because a fire and flood from one day tank room would potentially only affect the emergency diesel generator in the same train. Additionally, floodwater from the day tank room would be accommodated by floor drains in the diesel generator room and via a flow path through an exterior door to the outside of the diesel generator building. However, an unresolved question remained as to whether the as-built configuration of each day tank room met fire protection program requirements as described in the current licensing basis.
Licensing Basis: Columbias fire protection License Condition 2.C(14) states, in part, the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in Section 9.5.1 and Appendix F of the FSAR for the facility through Amendment 39. The License Condition also refers to letters submitted to the NRC and additional safety evaluations that approved the licensees fire protection program. Appendix F of the FSAR states that the diesel generator day tanks have 3000-gal capacity, the enclosures are capable of containing the entire contents of the day tanks, and there are no floor drains provided in the day tank rooms. However, the FSAR does not state that the day tank enclosures are also capable of containing any feasible amount of firewater. The NRCs safety evaluations documented approval of the licensees fire protection program, including deviations from BTP CMEB 9.5-1. Relevant deviations approved by the NRC involving the day tanks included installation of a 3000-gal tank and absence of floor drains in the enclosures.
After considering all available information, including consulting with fire protection experts in the NRC Office of Nuclear Reactor Regulation, the inspectors were unable to conclude whether the as-built configuration of the day tank enclosure was consistent with the current licensing basis. The inspectors determined that this issue met the criteria for the VLSSIR process because:
1. The inspection staff has not been able to conclude that the issue of concern is a
violation after considering licensee-provided supporting information and relevant information developed during the inspection process.
2. The specific issue of concern cannot have any potential to be Greater than Green
safety significance.
3. The resources required to resolve the current licensing basis question would not
effectively serve the agencys mission.
Significance: For the purpose of the VLSSIR process, the inspectors screened the issue of concern through IMC 0609, Appendix F, discussed the issue with a senior reactor analyst, and determined the issue of concern would likely be Green had a performance deficiency been identified.
Technical Assistance Request: No technical assistance request was processed for this issue.
Corrective Action Reference: CR 467070, CR 467903, CR 468153, CR 468453 Observation: Adverse Trend in QA/QC Program 71152S The inspectors identified an adverse trend in the QC/QA program based on review of condition reports that indicated the licensee failures in the following areas:
- (1) QC/QA personnel did not review work order packages for QC hold points prior to issuance of the work package;
- (2) QC inspectors did not fully understand the acceptance requirements for what they were inspecting;
- (3) workers failed to call QC inspectors to the field when the work order required a QC hold point; and
- (4) QC not being included on review of work being performed, resulting in the station inappropriately performing the work using a work request instead of a formal work order.
The QC/QA program is used to ensure that critical steps are completed correctly to ensure the systems can function when called upon. In the above highlighted examples, the station did not ensure the QC/QA program elements were implemented appropriately, indicating an adverse trend in the station's understanding of the QA/QC program requirements and importance. The station documented the adverse trend in CR 470350; however, the station continued to have the same issues as the above examples as indicated in the corrective action program.
Condition Reports: 468531, 468802, 468846, 468859, 468985, 469295, 469389, 470067, 470135, 470214, 470343, 470350, 470457, 470845, 470858, 470888, 471060, 471845, 471897, 471928,
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On May 20, 2025, the inspectors presented the emergency plan revisions in-office review results (Accession Number ML25141A024) inspection results to Steve English, Manager, Emergency Preparedness, and other members of the licensee staff.
- On May 21, 2025, the inspectors presented the inservice inspection results to Dave Brown, Site Vice President, and other members of the licensee staff.
- On June 4, 2025, the inspectors presented the radiation safety inspection results to Dave Brown, Site Vice President, and other members of the licensee staff.
- On July 29, 2025, the inspectors presented the integrated inspection results to Reginald Wainwright, Plant General Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Resulting from
Inspection
470795, 470796
Drawings
M544
Flow Diagram HVAC-Standby Gas Treatment Reactor
Building
Procedures
SOP-HPCS-LU
HPCS Valve and Breaker Lineup
Procedures
SOP-RCIC-LU
RCIC Valve and Breaker Lineup
Procedures
SOP-SCW-LU
Stator Coil Cooling System Valve and Breaker Lineup
Fire Plans
CGS PRE-FIRE
PLAN Turbine
Generator 471'
Fire Area: TG-1
Fire Plans
Radwaste 501-507
Fire Plans
R-2/U
Fire Area Reactor 501' Containment
Corrective Action
Documents
AR-
338347, 338394, 364053, 381703, 428932, 445323, 445643,
449125, 456450, 460275, 460501, 463112, 464567, 468526,
468613, 468997
Corrective Action
Documents
Resulting from
Inspection
CR-
469158
Drawings
M530-1
Flow Diagram Nuclear Boiler Recirculation System
Drawings
RRC-1550-4
3/4" Vent from RRC-V-67B to EDR
Procedures
ASME-P1-GTAW-
Gas Tungsten Arc Welding (GTAW) of Carbon Steels
Procedures
ASME-P1-
SMAW-1
Procedures
ASME-P8-GTAW-
GTAW Welding of Austenitic Stainless Steels
Procedures
PPM 8.3.425
Performance Demonstration Initiative Generic Procedures - ISI
Procedures
SPS-10-4
Guidance for Ultrasonic Examination Hands-on Practice
Procedures
SPS-3-3
Liquid Penetrant Examination - Columbia Generating Station -
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
SPS-4-3
Magnetic Particle Examination Columbia Generating Station -
Procedures
SPS-7-3
Visual Examination - Component Supports
Procedures
SPS-7-4
Visual Examination of Containment
Work Orders
WO 2139802, 2180311, 2167077, 2181697, 2190692, 218341201,
21349601, 221353401, 221357801, 221398801
Miscellaneous
R-27 Outage Shutdown Safety Plan
Procedures
1.3.83
Protected Equipment Program
Corrective Action
Documents
456234, 467464, 468661
Engineering
Changes
19603
Engineering
Changes
20698
Install temporary charger to power E-DP-S1/7
Work Orders
WO 210427, 2231104, 2231142
Corrective Action
Documents
469253
Procedures
10.25.132
Thrust Adjustments and Diagnostic Analysis of Motor
Operated Valves
Procedures
10.25.4
Lubrication and Inspection of Limitorque MOV9s)
Procedures
18.1.42
ASD Uncoupled Recirc Pump Motor Modification
Procedures
ESP-
R5051lMRRA-
B301
6.9KV Circuit Breaker RRC-CB-RRA Protective Relay - CC
Procedures
OSP-RCIC/IST-
C701
RCIC Post Maintenance Testing Prior to 150 PSIG
Procedures
OSP-RHR/IST-
R704
RHR Valve Operability-Refueling Shutdown
Procedures
RRC Pump Start Temperatures and Loop Flow Verification
Procedures
TSP-
ATWS/RPTA-
B501
ATWS Recirculation Pump A Trip System-LSFT
Procedures
TSP-MSIV-B801
Main Steam Isolation Valve Leak Rate Testing
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Work Orders
WO 206859, 2209154, 2209178, 2209463, 2210591, 2213357,
20819, 2233863
Corrective Action
Documents
Action Requests
(AR)
448618, 456539
Corrective Action
Documents
Resulting from
Inspection
Action Requests
(AR)
470162
Miscellaneous
50.54(q) Log #:
25-01
50.54(q) Screening and Effectiveness Evaluation - Replacing
Staffing of Augmenting and Support ERO Position Categories
with Pooled Assignment Approach
2/26/2025
Miscellaneous
50.54(q) Log #:
25-02
50.54(q) Screening and Effectiveness Evaluation -
Discontinuation of Use - ERO Notification Off-Site Pagers
03/10/2025
Miscellaneous
50.54(q) Log #:
25-03
50.54(q) Screening - Revision to EPIP PPM 13.1.1A,
Classifying the Emergency - Technical Bases, Revision 35
03/10/2025
Miscellaneous
Electronic Mail
from Steve
English, Columbia
Generating
Station, to Sean
Hedger, Nuclear
Regulatory
Commission
RE: RE: RE: Emergency Plan Revisions - Initial Review -
Requests
05/20/2025
Miscellaneous
Columbia Generating Station, Docket No. 50-397; Summary of
Changes and Analysis for Revision 69 of the EP-01 Columbia
Generating Station Emergency Plan
04/09/2025
Miscellaneous
Columbia Generating Station, Docket No. 50-397; Summary of
Changes and Analysis for Revision 46 of Plant Procedures
Manual 13.4.1 Emergency Notifications
04/24/2025
Miscellaneous
Columbia Generating Station, Docket No. 50-397; Summary of
Changes and Analysis for Revision 35 of Plant Procedures
Manual 13.1.1A Classifying the Emergency - Technical Bases
05/14/2025
Procedures
13.1.1A
Classifying the Emergency - Technical Bases
Procedures
13.4.1
Emergency Notifications
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
EPI-16
50.54(q) Change Evaluation
Procedures
SWP-EP-01
Emergency Response Organization and Training
Corrective Action
Documents
Action Request
(AR)
460750, 461159, 461351, 461770, 461832, 462162, 462203,
2963, 464195, 464908, 464915, 465483, 465501, 465771,
465915, 465938, 466272, 466342, 467123, 467153, 467563,
467581, 467622, 468380, 468403, 468828, 468979, 469068,
469076, 469214, 469218
Corrective Action
Documents
Resulting from
Inspection
Action Request
(AR)
469181
Engineering
Evaluations
2155
TEDE-ALARA Evaluation: Rapid Reassembly of RHR-V-8,
Entry Inside Valve Body
01/22/2025
Miscellaneous
1552.14
Energy Northwest Master Key Inventory Sheet
04/22/2025
Miscellaneous
DIC 1541.12
Energy Northwest Weekly LHRA and VHRA Door Checks
2/11/2025
Miscellaneous
PMID 23041
Spent Fuel Pool Material Inventory
Procedures
GEN-RPP-01
ALARA Program Description
Procedures
GEN-RPP-04
Entry Into, Conduct In, and Exit from Radiologically Controlled
Areas
Procedures
HPI-0.19
Radiation Protection Standards and Expectations
Procedures
PPM 1.11.15
Control of Radioactive Material
Procedures
PPM 11.2.13.1
Radiation and Contamination Surveys
Procedures
PPM 11.2.13.8
Airborne Radioactivity Surveys
Procedures
PPM 11.2.15.13
Control of Personnel Skin and Clothing Contamination
Procedures
PPM 11.2.15.7
Release of Material from Radiologically Controlled Areas
Procedures
PPM 11.2.2.14
Radiological Planning and Reviews
Procedures
PPM 11.2.7.1
Area Posting
Procedures
PPM 11.2.7.3
High Radiation Area, Locked High Radiation Area, and Very
High Radiation Area Controls
Procedures
PPM 11.2.8.2
Radiation Work Permit Preparation and Use
Procedures
SWP-RPP-01
Radiation Protection Program
Radiation Surveys M-20250205-3
Radwaste 437 Floor All Areas
2/05/2025
Radiation Surveys M-20250327-12
Reactor Building 572 Monthly
03/27/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Radiation Surveys M-20250327-2
Reactor Building 501 Floor
03/27/2025
Radiation Surveys M-20250403-2
Reactor Building 441 and 422 North Floor
04/03/2025
Radiation Surveys M-20250403-9
Reactor Building 548
04/04/2025
Radiation Surveys M-20250406-1
Radwaste 467 All Areas
04/06/2025
Radiation Surveys M-20250410-10
Radwaste 507 All Areas
04/10/2025
Radiation Surveys M-20250410-11
Reactor Building 606 Floor
04/10/2025
Radiation Surveys M-20250416-8
Reactor Building 471 Floor
04/16/2025
Radiation Surveys M-20250417-30
Reactor Building 522 Floor Map
04/17/2025
Radiation Surveys M-20250418-33
Radwaste 437 Floor All Areas
04/18/2025
Radiation Surveys M-20250419-31
R27 RHR-V-8 Breach
04/19/2025
Radiation Work
Permits (RWPs)
30005118
R27 Control Rod Drive Mechanisms (CRDM) U/V Remove and
Replace 20 CRDMs
Radiation Work
Permits (RWPs)
30005128
R27 *LHRA* Drywell NRC Tours
Radiation Work
Permits (RWPs)
30005137
R27 Drywell/Steam Tunnel *LHRA* ISI/Scaffold/Insulation
Radiation Work
Permits (RWPs)
30005140
R27 CRDM Positioner Indicator Probe (PIP) Replacements
Radiation Work
Permits (RWPs)
30005276
R27 Drywell DCA/VR RHR-V-8 Replace Wedge and Seat
HR/MP
Self-Assessments
AR-SA 00453390-
Annual Review of Columbia Generating Station Radiation
Protection Program (RPP) to meet 10CFR20.1101(c)
Requirements
04/10/2025
Self-Assessments
AR-SA 00467622
Snapshot Self-Assessment Report: IP 71124, Attachment 1,
"Radiological Hazard Assessment and Exposure Controls"
2/28/2025
Corrective Action
Documents
Action Request
(AR)
27385, 445677, 445797, 449789, 450111, 456803, 456820,
469319, 469351
Corrective Action
Documents
Resulting from
Inspection
Action Request
(AR)
469181, 469292, 469319, 469326, 469351
Procedures
GEN-RPP-05
Respiratory Protection Program Description
Procedures
GEN-RPP-10
Use of Respiratory Protection Equipment
Procedures
MSP-WMA-B101
Control Room DIV A Emergency Filtration System HEPA Filter
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Test
Procedures
MSP-WMA-B102
Control Room DIV B Emergency Filtration System HEPA Filter
Test
Procedures
MSP-WMA-B103
Control Room DIV A Filtration System - Carbon Adsorber Test
Procedures
MSP-WMA-B104
Control Room DIV B Filtration System - Carbon Adsorber Test
Procedures
PPM 10.2.62
Breathing Air Compressor Operation
Procedures
PPM 10.2.82
HEPA Filter In-Place Testing
Procedures
PPM 10.2.83
Carbon Filter In-Place Testing
Procedures
PPM 10.27.49
Reactor Building In-Plant Particulate Monitor RRA-RIS-3
Procedures
PPM 10.27.50
Radwaste Building In-Plant Particulate Monitor WRA-RIS-1
Procedures
PPM 10.27.51
Radwaste Building In-Plant Particulate Monitor WRA-RIS-2
Procedures
PPM 11.2.10.17
Procedures
PPM 11.2.13.11
Characterization of Alpha Radioactivity
Procedures
PPM 11.2.15.11
Use and Certification of Portable Air Handling Units
Procedures
PPM 11.2.15.13
Control of Personnel Skin and Clothing Contamination
Procedures
PPM 11.2.9.41
AMS-4 Continuous Air Monitor (CAM)
Procedures
PPM 13.14.4
Emergency Equipment Maintenance and Testing
Radiation Surveys M-20230830-3
RW 437 NUPAC Cage Replace Title: SERDS filter
08/30/2023
Radiation Work
Permits (RWPs)
30005021
23 RW 437 **LHRA** Waste Processing NUPAC Cage
Radiation Work
Permits (RWPs)
30005058
23 RX 606 *HR**HCA* FHB Maintenance and Floor Work
Radiation Work
Permits (RWPs)
30005114
25 PRE/POST R27 RX 606 HRA
Radiation Work
Permits (RWPs)
30005115
25 PRE/POST R27 HRA
Radiation Work
Permits (RWPs)
30005239
24 RW 487 WEA FAN REM/REPLACE PRE-FILTER HRA /
ARA
Radiation Work
Permits (RWPs)
30005273
24 PRE R27 RX 606 HRA
Radiation Work
Permits (RWPs)
30005339
25 RX 501 CRD REBUILD RM WORK HRA
Self-Assessments
AU-CH-24
Chemistry and Environmental Monitoring Programs
10/24/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Self-Assessments
AU-RP/RW-24
Radiation Protection and Process Control Programs
01/15/2025
Work Orders
2175691 01
Control Room Division B Filtration System - Charcoal
Adsorber Test
2/05/2023
Work Orders
2175692 01
Control Room DIV B Emergency Filtration System HEPA Filter
Test
2/02/2023
Work Orders
2189196 01
Sample Plant Breathing Air
01/23/2023
Work Orders
2189197 01
Sample Plant Breathing Air
05/05/2023
Work Orders
200049 01
Control Room Division A Filtration System - Carbon Adsorber
Test
04/14/2025
Work Orders
204967 01
Sample Plant Breathing Air
07/26/2023
Work Orders
204968 01
Sample Plant Breathing Air
10/23/2023
Work Orders
204976 01
Sample Plant Breathing Air
03/24/2024
Work Orders
204977 01
Sample Plant Breathing Air
06/24/2024
Work Orders
216332 02
Fire Brigade Station Inventory Station 1 TG 441
04/02/2024
Work Orders
219230 01
Sample Plant Breathing Air
09/16/2024
Work Orders
219231 01
Sample Plant Breathing Air
2/03/2024
Work Orders
219273 01
Sample Plant Breathing Air
03/16/2025
71151
Self-Assessments
AR-SA 00464361
Snapshot Self-Assessment Report: PI - 71151 - OR01 PI
Verification: Occupational Exposure
03/12/2025
71151
Self-Assessments
AR-SA 00464391
Snapshot Self-Assessment Report: PI Verification PR01:
Radiological Effluent Technical Specifications/Offsite Dose
Calculation Manual/Radiological Effluent Occurrences
01/27/2025
Calculations
ME-02-02-40
PFSS Flooding Analysis - Diesel Generator Building
Calculations
ME-02-03-02
Calculation for Diesel Generator Building Flooding Analysis
Corrective Action
Documents
467070, 467903, 468153, 468453,
Engineering
Evaluations
TM-2210
Flooding Review - Impact to Adjacent Areas
Fire Plans
CGS Pre-Fire Plan
Miscellaneous
Columbia Generating Station Final Safety Analysis Report
Corrective Action
Documents
461519, 465452, 466613, 467084, 467154, 469058, 469119