IR 05000454/2025002
| ML25225A028 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 08/13/2025 |
| From: | Dariusz Szwarc NRC/RGN-III/DORS/RPB3 |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| IR 2025002 | |
| Download: ML25225A028 (1) | |
Text
SUBJECT:
BYRON STATION - INTEGRATED INSPECTION REPORT 05000454/2025002 AND 05000455/2025002
Dear David Rhoades:
On June 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Byron Station. On July 7, 2025, the NRC inspectors discussed the results of this inspection with Harris Welt, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at Byron 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 III; and the NRC Resident Inspector at Byron Station.
August 13, 2025 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, Dariusz Szwarc, Chief Reactor Projects Branch 3 Division of Operating Reactor Safety Docket Nos. 05000454 and 05000455 License Nos. NPF-37 and NPF-66 Enclosure:
As stated cc w/ encl: Distribution via LISTSERV Signed by Szwarc, Dariusz on 08/13/25
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Byron Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Develop, Implement and Maintain FLEX Strategies Consistent with the Requirements of 10 CFR 50.155(b)(1)
Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000454,05000455/2025002-01 Open/Closed
[H.8] -
Procedure Adherence 71111.20 An NRC-identified Green finding and associated non-cited violation (NCV) of 10 CFR 50.155,
Mitigation of Beyond-Design-Basis Events, was identified for the licensees failure to develop, implement, and maintain the equipment relied upon for the sites mitigation strategies for beyond-design-basis external events (BDBEE), typically referred to as FLEX strategies.
Specifically, the station failed to perform the required pre-deployment of FLEX equipment and/or other contingency plan actions to assure FLEX strategies remain valid during outage conditions as outlined in the Refueling Outage B2R25 Shutdown Safety Contingency Plan and Work Order 5569966, Byron Unit 2 Preoutage Tasks. As a result, considering maximum debris removal time, the Phase 2 FLEX strategy would not have been able to be implemented prior to exceeding the time to top-of-active fuel (TAF) and start of core damage.
Exceeding Core Operating Limits Report Minimum Pressure Limits due to Isolated Steam Dumps During Power Ascension Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000455/2025002-02 Open/Closed
[H.4] -
Teamwork 71111.20 A self-revealed finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Technical Specifications (TS) 5.4.1 Procedures, was identified when the licensee failed to implement a procedure for power ascension when exiting a refueling outage.
Specifically, the licensee failed to ensure that manual isolation valves for the steam dumps were open prior to raising reactor power above 5 percent and entering Mode 1 as required by Procedure 2BGP 100-3, Revision 110, Power Ascension. The resulting transient caused the 2B steam generator (SG) power operated relief valve (PORV) to lift, and pressurizer pressure to drop below the Core Operating Limits Report (COLR) minimum pressure limit of 2209 pounds per square inch gauge (PSIG).
Additional Tracking Items
None.
PLANT STATUS
Unit 1 operated at or near rated thermal power for the entire inspection period. Unit 2 began the inspection period operating at full achievable power until the unit was shut down on April 14, 2025, for refueling outage (RFO) B2R25. Following completion of the refueling outage, the reactor was restarted and taken critical on May 1, 2025. The unit reached full power operation on May 5, 2025, and continued operating at or near full 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
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated station readiness prior to the onset of seasonal hot temperatures for various safety related systems and pieces of equipment such as, auxiliary feedwater, chilled water systems, primary water systems, and rod control, to mention only a couple, needed for reliable operation during the summer period.
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)partial walkdown of Unit 2 decay heat removal support systems during periods of increased shutdown risk ending on April 20, 2025 (2)partial walkdown of the 1B emergency diesel generator (EDG) prior to 6-year maintenance window of the 1A EDG ending on May 23, 2025
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
(1)complete walkdown of the Unit 2 safety injection system ending on April 30, 2025
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Fire Zone 18.11-0, elevation 686' & 702', river screen house on April 9, 2025
- (2) Fire Zone 11.4A-2, auxiliary building, elevation 383-0, Unit 2 2B auxiliary diesel feedwater pump and day tank room during hot work on April 18, 2025
- (3) Fire Zone 8.5-2, turbine building, elevation 426', Unit 2 general areas during various RFO B2R25 welding projects ending April 21, 2025
- (4) Fire Zone 18.3-2, miscellaneous area, elevation 377'-0," main steam and auxiliary feedwater pipe tunnel on April 22, 2025
- (5) Fire Zone 8.3-1, Unit 2 turbine building, elevation 401'-0," northwest, following failed fire protection deluge testing of the 241 UAT, as documented in Action Request (AR) 4861179 on April 28, 2025
- (6) Fire Zone 11.1A-0 and 11.1B-0, auxiliary building, elevation 330'-0," Unit 1 and 2 essential service water pump room on June 25, 2025
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors reviewed internal flooding mitigation protections in the Unit 2 auxiliary building due to essential service water system moderate energy line breaks during planned removal of flood seals during the week of April 7, 2024.
===71111.08P - Inservice Inspection Activities (PWR) The inspectors assessed the licensees degradation monitoring of the reactor coolant system boundary, reactor vessel internals, risk significant piping system boundaries, and containment boundary, as well as the fabrication, examination and acceptance of repair and replacement activities by reviewing the following in Unit 2 during RFO B2R25 from April 14, 2025, to April 24, 2025.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===
The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:
- (1) Ultrasonic Examination
- Reactor Coolant System Piping Welds, 2RC28A-3 J03-5, Cat. R-A, Item R1.11-2
- 24 Loop Stop Isolation Valve Studs, 2RC8001A, Cat. B-G-1, Item B6.210 Magnetic Particle Examination
- Reactor Pressure Vessel (RPV) Bottom Head Visual Exam of Bottom Mounted Instrument Penetrations and Bare Metal Surface, Code Case N-722-1, Item B15.80
- VT-1 of 2AF-01-PA/E-1 Welded Aux Feed Pump Attachment, Cat. D-A, Item D1.30
- VT-3 of 2SI05DA-6/2SI01009X Pipe Strut, Cat. F-A, Item F1.10B
- VT-3 of 2SI08FA-3/2SI23009X Pipe Strut, Cat. F-A, Item F1.10B Welding Activities
Weld Repair of Through Wall Flaw on Unit 2 Essential Service Water Valve 2SX035, Work Order (WO) 5416744 Base Metal Weld Repair of 2A CS Cubicle Cooler Tubesheet and Divider Plate, WO 5520062 PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection
Activities (IP Section 03.02) (1 Sample)
The inspectors reviewed the licensees inspection of the following vessel upper head penetrations:
- (1) Bare Metal Visual Examination of RPV Upper Head Surface Penetrations PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)
The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:
- (1) Boric Acid Evaluations and Corrective Actions
1. Boric Acid Leakage on Loop Stop Isolation Valve 2RC8001C, AR 4708057
2. Boric Acid Leakage on U2 Boric Acid Pump Skid, AR 4790515
3. Boric Acid Leakage on 2CV206, AR 4815953
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 various activities involving on-watch operations crews during the week of June 16, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated performance of a complex casualty graded scenario by a crew of licensed plant operators in the facilitys simulator on May 20, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples 1 Partial)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1)review of evaluations of maintenance rule condition monitoring events for NAL cards following failure of 1LT-0548 1PA03J card on December 20, 2024 (2)maintenance effectiveness review of fuel injector replacement in the 1B auxiliary feedwater pump (3)
(Partial)maintenance effectiveness review of the 0A essential service water makeup pump
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:
(1)maintenance effectiveness, performance history, and quality reviews for conditionally released components (Limitorque Torque Switch P/N 11501-010, FRA P/N 1167723-662) affected by the Framatome Part 21 notification
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)review and evaluation of the risk associated with the failure to properly stage FLEX equipment used for shutdown contingency plans, as documented in AR 4857767 (2)review and evaluation of the risk associated with various troubleshooting activities on Unit 1 and Unit 2 main steam isolation valves (MSIVs) following lowered electrohydraulic (EHC) fluid levels (3)review and evaluation of the risk associated with 125 Vdc bus 112 ground troubleshooting, as documented in AR 4876304
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
(1)evaluation of the operability of the Unit 1 and 2 SX016A/B, reactor containment fan cooler (RCFC) essential service water (SX) supply isolation valves, and SX027A/B, RCFC SX return header isolation valves, following issues and concerns with diagnostic testing, as documented in ARs 4804981 and 4804996 (2)evaluation of the operability of the Unit 1 auxiliary feedwater system following issues with stroke time testing of the 1AF005E, 1B auxiliary feedwater pump discharge to 1A steam generator flow control valve, as documented in AR 4843419 (3)evaluation of the operability of the full flow capability of the Chemical and Volume control (CV) and safety injection systems following discovery of flow transmitters being out of tolerance, as documented in ARs 4858648 and 4858587 (4)evaluation of the operability of the ultimate heat sink following identification of corrosion on 0SX163G, 0B SX cooling tower riser isolation valve, as documented in AR 4871784 (5)evaluation of the operability of division 211 125 Vdc battery, as documented in AR 4868814
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated RFO B2R25 activities from April 4, 2025, to May 2, 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) (9 Samples)
(1)functional and operational testing of the 2A residual heat removal (RHR) pump suction valves following a scheduled work window, as documented in WO 5570096 (2)functional and operational testing of 2AF005B, 2B auxiliary feedwater pump discharge to 2B steam generator flow control valve, following valve being stuck in midposition during testing, as documented in AR 4853774 (3)functional and operational testing of the 2A EDG following replacement of the electronic governor, as documented in WO 5410275 (4)functional and operational testing of the 2A EDG during sequencing testing following scheduled maintenance during RFO B2R25, as documented in WO 5410997 (5)functional and operational testing of the Unit 2 reactor core through low power physics testing following RFO B2R25, as documented in WO 5413846 (6)functional and operational testing of the 2A essential service water pump following pump motor replacement, as documented in WO 5413158 (7)functional and operational testing of Unit 2 main steam safety valves following replacement during RFO B2R25, as documented in various work orders (8)functional and operational testing of the 1A EDG following a 6-year PM work window, as documented in various work orders (9)functional and operational testing of 2RE1003, reactor coolant drain tank (RCDT)pump discharge containment isolation valve, following diaphragm replacement, as documented in WO 5637172
Surveillance Testing (IP Section 03.01) (3 Samples)
(1)2BVSR 7.1.1-1, Unit 2 Main Steam Safety Valves Operability Test, as documented in various work orders (2)2BOSR 3.1.5-1, Unit Two Train A Solid State Protection System Surveillance, as documented in WO 5411972 (3)1BOSR 8.1.2-2, Unit One 1B Diesel Generator Operability Surveillance, as documented in WO 5599755
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)0BOSR 5.5.8.SX.5-2c, Unit 0 Comprehensive Inservice Testing (IST) Requirements for Essential Service Water Makeup Pump 0B, as documented in WO 5630793
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
(1)local leak rate testing of primary containment isolation valves throughout RFO B2R25, as documented in various work orders
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
The inspectors evaluated:
- (1) The inspectors evaluated the sites emergency plan by observing emergency response organization health physics drill on May 14,
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) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
(1)licensee surveys of workers and potentially contaminated material leaving the radiologically controlled area (RCA) at the auxiliary building during a refueling outage (2)licensee surveys of workers and potentially contaminated material leaving the RCA at the Unit 2 containment access facility during a refueling outage
Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)
The inspectors evaluated the licensees control of radiological hazards for the following radiological work:
- (1) Unit 2 reactor head disassembly
- (2) Unit 2 reactor head and upper internals move
- (3) Unit 2 seal table activities
- (4) Unit 2 reactor cavity decontamination High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (3 Samples)
The inspectors evaluated licensee controls of the following high radiation areas and very high radiation areas:
(1)spent resin transfer pump room
- (2) Unit 2 incore detector pit (3)0B auxiliary equipment drain tank 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
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) potable HEPA unit 110354 maintaining negative ventilation under the Unit 2 reactor head
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) Unit 1 (April 1, 2024, through March 31, 2025)
- (2) Unit 2 (April 1, 2024, through March 31, 2025)
MS06: Emergency AC Power Systems (IP Section 02.05) (2 Samples)
- (1) Unit 1 (April 1, 2024, through March 31, 2025)
- (2) Unit 2 (April 1, 2024, through March 31, 2025)
MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)
- (1) Unit 1 (April 1, 2024, through March 31, 2025)
- (2) Unit 2 (April 1, 2024, through March 31, 2025)71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03) (1 Sample 1 Partial)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) The inspectors reviewed the corrective actions associated with the use of an outdated procedure during main steam safety valve testing, as documented in AR 4855906.
(2)
(Partial)
The inspectors reviewed the implementation and tracking of the essential service water fouling monitoring program (BVP 800-30) and extent-of-condition exams.
INSPECTION RESULTS
Failure to Develop, Implement and Maintain FLEX Strategies Consistent with the Requirements of 10 CFR 50.155(b)(1)
Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000454,05000455/2025002-01 Open/Closed
[H.8] -
Procedure Adherence 71111.20 An NRC-identified Green finding and associated non-cited violation (NCV) of 10 CFR 50.155, Mitigation of Beyond-Design-Basis Events, was identified for the licensees failure to develop, implement, and maintain the equipment relied upon for the sites mitigation strategies for beyond-design-basis external events (BDBEE), typically referred to as FLEX strategies. Specifically, the station failed to perform the required pre-deployment of FLEX equipment and/or other contingency plan actions to assure FLEX strategies remain valid during outage conditions as outlined in the Refueling Outage B2R25 Shutdown Safety Contingency Plan and WO 5569966, Byron Unit 2 Preoutage Tasks. As a result, considering maximum debris removal time, the Phase 2 FLEX strategy would not have been able to be implemented prior to exceeding the time to top-of-active fuel (TAF) and start of core damage.
Description:
Following the events at the Fukushima Daiichi Nuclear Power Plant in Japan, the NRC issued Orders (e.g., EA-12-049) which required each nuclear power plant licensee to develop strategies for mitigating a BDBEE like the one at Fukushima Daiichi. Title 10 CFR 50.155 codified many of the post-Fukushima orders. In particular, Section 50.155(b), Strategies and Guidelines, Section (1), requires, in part, licensees to develop, implement, and maintain strategies and guidelines for mitigating a BDBEE.
Procedure CC-AA-118, Diverse and Flexible Coping Strategies (FLEX), Spent Fuel Pool Instrumentation (SFPI), and Hardened Containment Vent System (HCVS) Program Document, is the Constellation fleet procedure that provides governance and procedural guidelines for BDBEEs and defines the sites compliance with 10 CFR 50.155 and NRC Orders.
Procedure CC-BY-118-1004, Byron Station Units 1 & 2 Final Integrated Plan Document Mitigation Strategies for a Beyond-Design-Basis External Event, implements the fleet program by outlining site-specific details and provides Byrons approach to comply.
Section 2.7, Shutdown and Refueling Mode, assumes during an extended loss of AP power (ELAP) when in cold shutdown or refueling modes, installed plant systems cannot be relied upon, and transition to Phase 2 will begin immediately. Attachment 8, Sequence of Events Timeline, presents the timeline for an ELAP/loss of ultimate heat sink (LUHS) event at Byron. Per this attachment, the time to clear debris to allow equipment deployment is assumed to be up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Procedure CC-BY-118-1003, Byron FLEX (BDBEE) Validation Process, outlines the process to validate that FLEX strategies are feasible and may be executed within the necessary time requirements. Attachment 2, Validation Plan 10, describes the validation times for staging and connecting Phase 2 FLEX pumps. The time required to deploy and stage the FLEX pump suction and discharge hoses is listed as 25 minutes. The time required to connect the hoses is listed as 12 minutes. This time, combined with the time required to deploy, stage, and hook up the FLEX pump hoses, would require a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 37 minutes.
Procedure OU-AA-103, Shutdown Safety Management Program, provides guidance to determine if outage conditions and activities require pre-deployment of FLEX equipment and/or other contingency plan actions to assure FLEX strategies will remain valid.
4, Shutdown/Refueling Modes FLEX Capability Assessment Process, requires that, during periods of time when the TAF is less than the time to deploy the FLEX equipment with assumed maximum debris removal time, pre-deployment of FLEX equipment is required.
The RFO B2R25 shutdown safety plan provides Contingency Plan #2 to provide actions to ensure the FLEX strategy is valid during short TAF periods. The plan states the shortest TAF is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 2 minutes. This occurs during the initial drain down to disassemble the reactor head.
On April 9, per the WO 5569966 task 43 sub-task 4 Pre-stage FLEX equipment per Outage Shutdown Safety was noted as done.
On April 14, 2025, Byron Station Unit 2 entered RFO B2R25.
On April 15, the unit lowered pressurizer level to support reactor head removal. At 1237 on April 15, the unit shutdown risk was declared Yellow due to the lowered reactor coolant inventory and high decay heat load.
On April 16 at 1248, the shutdown risk returned to Green following restoration of reactor coolant inventory. This period of approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> encompasses the period of time when TAF was 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 2 minutes. In the evening, while performing plant tours, the resident inspectors identified that FLEX hoses necessary for implementation of the Phase 2 FLEX strategy had not been staged as required per the Shutdown Safety Contingency Plan.
As a result, during the period of time when TAF was shorter than the time required to deploy the Phase 2 FLEX equipment, the station would not have been able to implement the FLEX strategy prior to exceeding TAF and the start of core damage.
Corrective Actions: The licensee staged the required hoses and trailer. Additional actions were created to better clarify the work order task instructions regarding the staging of equipment.
Corrective Action References: AR 4857767, NRC Identified - FLEX Pump Staged Without Hoses
Performance Assessment:
Performance Deficiency: The inspectors determined the licensees failure to pre-deploy the FLEX hoses in support of implementing Phase 2 FLEX strategies during the period of time when TAF was less than the time to deploy the FLEX equipment such that FLEX strategies remained valid consistent with 10 CFR 50.155 was a performance deficiency. Specifically, the station failed to perform the required pre-deployment of FLEX equipment and/or other contingency plan actions to ensure FLEX strategies remain valid during outage conditions as outlined in the RFO B2R25 Shutdown Safety Contingency Plan and WO 5569966.
As a result, considering maximum debris removal time, the Phase 2 FLEX strategy would not have been able to be implemented prior to exceeding the TAF and start of core damage.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to pre-deploy all necessary FLEX equipment which would have adversely affected the availability, reliability, and capability of the FLEX equipment to respond to initiating events to prevent undesirable consequences during the period of time when TAF was less than the time to deploy the FLEX equipment.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. The inspectors screened the finding using 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 3, and answered no to all questions through B.8, and therefore, the finding screens to very low safety significance (Green).
Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, WO 5569966 states in the instructions to pre-stage the FLEX equipment per Contingency Plan #2 of the Outage Shutdown Safety Plan. The shutdown safety plan specifies this to be the medium-head FLEX pump and associated hoses. Despite being marked as complete on April 9, 2025, the hoses had not been pre-staged per the WO instructions until after April 16, 2025.
Enforcement:
Violation: 10 CFR 50.155(b)(1), states, in part, that each applicant or licensee shall develop, implement, and maintain strategies and guidelines to mitigate beyond-design-basis external events from natural phenomena that are developed assuming a loss of all ac power concurrent with a loss of normal access to the ultimate heat sink.
Procedure CC-BY-118-1004, Byron Station Units 1 & 2 Final Integrated Plan Document Mitigation Strategies for a Beyond-Design-Basis External Event, implements the fleet program by outlining site-specific details and provides Byrons approach to comply.
Section 2.7, Shutdown and Refueling Mode, assumes during an ELAP when in cold shutdown or refueling modes, installed plant systems cannot be relied upon and transition to Phase 2, the documents states When in Cold Shutdown and Refueling, many variables exist which impact the ability to cool the core. In the event of an ELAP during these Modes, installed plant systems cannot be relied upon to cool the core, thus transition to Phase 2 will begin immediately. All efforts will be made to expeditiously provide core cooling and minimize heat-up and repressurizationAttachment 8, Sequence of Events Timeline, presents the timeline for an ELAP/LUHS event at Byron. Per the validation time Time to clear debris to allow equipment deployment is assumed to be up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This time is considered to be reasonable based on-site reviews of the deployment paths and the location of the FSRB.
Debris removal equipment is stored in the FSRB. This time, combined with the time required to deploy, stage, and hook up the FLEX pump hoses, would require a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 37 minutes. The Attachment 8 contains a table that elaborate this and other timelines that the licensee agreed to follow.
Procedure OU-AA-103, Shutdown Safety Management Program, provides guidance to determine if outage conditions and activities require pre-deployment of FLEX equipment and/or other contingency plan actions to assure FLEX strategies will remain valid.
4, Shutdown/Refueling Modes FLEX Capability Assessment Process, requires that, during periods of time when the TAF is less than the deployment time then Lower deployment time: - increase resources; - Assign FLEX actions; - Brief personnel; - Develop written contingency plan; Pre-deploy FLEX equipment to reduce deployment time < time to TAF. May be partial pre-deployment
WO 5569966, Byron Unit 2 Preoutage Tasks, was established, in part, to ensure that pre-staging of FLEX equipment was done prior to the start of Unit 2 RFO B2R25.
Contrary to the above, from 1237 on April 15 until 1248 on April 16, 2025, the licensee failed to implement the strategies and guidelines to mitigate BDBEE assuming a loss of all ac power concurrent with a loss of normal access to the ultimate heat sink, as detailed in CC-BY-118-1004, when they failed to pre-stage all required Phase 2 FLEX equipment.
Specifically, during the period of time when TAF was shorter than the time to deploy FLEX, the licensee failed to maintain the ability to implement their Phase 2 FLEX strategy due to the failure to fully execute WO 5569966 to pre-stage FLEX hoses prior to RFO B2R25.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Exceeding Core Operating Limits Report Minimum Pressure Limits due to Isolated Steam Dumps During Power Ascension Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000455/2025002-02 Open/Closed
[H.4] -
Teamwork 71111.20 A self-revealed finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Technical Specifications (TS) 5.4.1 Procedures, was identified when the licensee failed to implement a procedure for power ascension when exiting a refueling outage. Specifically, the licensee failed to ensure that manual isolation valves for the steam dumps were open prior to raising reactor power above 5 percent and entering Mode 1 as required by Procedure 2BGP 100-3, Revision 110, Power Ascension. The resulting transient caused the 2B steam generator (SG) power operated relief valve (PORV) to lift and pressurizer pressure to drop below the Core Operating Limits Report (COLR) minimum pressure limit of 2209 pounds per square inch gauge (PSIG).
Description:
On May 1, 2025, with Unit 2 in Mode 2 following the B2R25 refueling outage, the licensee raised reactor power to greater than 5 percent and entered Mode 1 prior to opening all manual isolation valves for steam dumps as required by Step F.8.c.2 of Procedure 2BGP 100-3, Revision 110, Power Ascension, the procedure containing the steps necessary to raise power from startup in Mode 2 to full operation in Mode 1. Nine steam dumps were isolated for low power physics testing and had not been restored when the direction to raise power was given. The remaining available steam dumps opened to maintain target steam pressure but due to continued power ascension, the 2B SG PORV opened to lower the increasing SG pressure.
The resulting transient caused pressurizer pressure to drop below the COLR minimum pressure limit of 2209 PSIG as required by TS Limiting Condition for Operation (LCO) 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits.
The Technical Specification Action Condition (TSAC) requires that, when pressurizer pressure drops below the minimum pressure required by the COLR, that pressurizer pressure be restored to within the COLR limit of 2209 PSIG within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The licensee entered TSAC 3.4.1.a at 1051, restored pressurizer pressure to within the COLR limit at 1052, and subsequently exited the TSAC.
Inspectors onsite on May 1, 2025, responded to the control room to monitor corrective actions taken by licensee personnel following the transient.
Corrective Actions: After identifying that nine of the available steam dumps were isolated, the licensee dispatched equipment operators to open the steam dump isolation valves.
The licensee also conducted a human performance review board as well as a prompt investigation into the cause of the event and performed a site standdown outlining the event.
Corrective Action References: AR 4862325, 2B SG PORV Lifted During Plant Power Ascension
Performance Assessment:
Performance Deficiency: The inspectors determined the licensees failure to implement Procedure 2BGP 100-3, Revision 110, was contrary to TS 5.4.1 Procedures and was a performance deficiency. Specifically, the licensee failed to perform Step F.8.c.2 of Procedure 2BGP 100-3 which requires the licensee to Verify/Open 2MS003A-M and 2MS005A-M, Manual Isolation valves for Steam Dump Valves prior to continuing to Step F.8.c.3 and raising power to 5 percent and entering Mode 1. The failure to perform Step F.8.c.2 resulted in a pressure transient and the opening of the 2B SG PORV, leading to exceeding the COLR for allowed pressurizer pressure.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to implement Procedure 2BGP 100-3 resulted in a configuration control error during startup that upset plant stability when nine of the available steam dumps were isolated and remained closed during power accension. The 2B SG PORV opened to maintain pressure, causing a plant transient that resulted in violating the COLR limit for pressurizer pressure.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
The inspectors determined that the finding was of very low safety significance (i.e., Green)because they answered No to the Transient Initiators Question in Exhibit 1 - Initiating Events Screening Questions. Specifically, the finding did not cause a reactor trip, and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, during the transition to Mode 1, the shift manager was not aware of the current status of all of the steam dumps. While some crew members may have been aware, this information was not adequately conveyed to the shift manager prior to the direction to raise power and the transition to Mode 1.
Enforcement:
Violation: Technical Specification Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
NRC Regulatory Guide 1.33, Revision 2, Appendix A, Section D, addresses Implementation and Section D.2 addresses General Plant Operating Procedures.
The licensee established Procedure 2BGP 100-3, Revision 110, to address power ascension, a general plant operating procedure. Procedure 2BGP 100-3, Revision 110, Step F.8.c.2 requires the licensee to verify/open 2MS003A-M and 2MS005A-M, Manual Isolation Valves for Steam Dump Valves.
Contrary to the above, on May 1, 2025, the licensee failed to implement Step F.8.c.2 of Procedure 2BGP 100-3, Revision 110. Specifically, the licensee failed to verify/open all valves outlined in Step F.8.c.2 before proceeding to Step F.8.c.3 and raising power to greater than or equal to 5 percent power. This resulted in the 2B SG PORV lifting and entry into TS 3.4.1 due to lowered pressurizer pressure below allowed COLR limits.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Incorrect Procedure Revision Used During Main Steam Safety Valve Testing 71152A The inspectors performed a detailed review of several ARs related to the performance of main steam safety valve (MSSV) testing on Unit 2 prior to the start of RFO B2R25 on April 10, 2025.
During field observations of the MSSV testing, the inspectors identified that the licensee was using Revision 2 of BMP 3114-21, Main Steam Safety Valve Testing Using Setpoint Verification Device, instead of Revision 5, the current revision. A work group evaluation performed under AR 4855906, NRC identified incorrect procedure revision used, concluded that work packages were pulled from the library and not validated to have the current revision of BMP 3114-21 contained within them. Further, the first line supervisors and mechanics performing the work failed to validate the correct revision prior to starting work. This is contrary to licensee Procedure HU-AA-104-101, Procedure Use and Adherence, which states procedure users use only the current revision of the procedure and was a performance deficiency. Three MSSVs were tested with the incorrect revision of the procedure before work was stopped, and the correct revision was implemented for use.
Investigations into the impact of the incorrect procedure revision revealed that Revision 2 did not require the verification of thermal equilibrium prior to the start of set pressure testing, which was required by the 2004 ASME OM Code, Section I-8110. Engineering Change 644188 evaluated the conditions of the three MSSVs prior to the start of testing and concluded that, though there were identified changes in temperature, the thermal equilibrium requirements from Section I-8110 of the ASME OM Code were met for all valves.
The inspectors performed a review of the licensee evaluation, the test conditions and the code requirements and determined the identified changes in temperature during the test would be unlikely to have a significant impact on the outcome of the test. Therefore, the inspectors determined this was a minor violation and no findings or violations of more than minor significance were identified during this review.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified that no proprietary information was retained or documented in this report.
- On July 7, 2025, the inspectors presented the integrated inspection results to Harris Welt, Site Vice President, and other members of the licensee staff.
- On April 24, 2025, the inspectors presented the inservice inspection results to Harris Welt, Site Vice President, and other members of the licensee staff.
- On April 25, 2025, the inspectors presented the radiation protection inspection results to Jacob Burton, Radiation Protection Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
Letter from Harris Welt, Byron Generating Station Site Vice
President, to John Kowalski, Senior Vice President of
Nuclear Operations Constellation Nuclear: Certification of
25 Summer Readiness
05/14/2025
OP-AA-108-111-
1001
Severe Weather and Natural Disaster Guidelines
Procedures
Seasonal Readiness
BOP DG-E1B
Unit 1 Diesel Generator Train B Electrical Lineup
BOP DG-M1B
Unit 1 Diesel Generator System Train B Valve Lineup
BOP RH-E2
Residual Heat Removal System Electrical Lineup
BOP RH-M2
Residual Heat Removal System Valve Lineup
BOP SI-E2
Unit 2 Safety Injection System Electrical Lineup
BOP SI-M2
Unit 2 Safety Injection System Valve Lineup
BOP SX-E2B
Unit 2 Essential Service Water Train B Electrical Lineup
Procedures
BOP SX-M2B
Unit 2 Essential Service Water Train B Valve Lineup
Corrective Action
Documents
Deluge Piping Failed - Sprayed 233Y - 2BOA ELEC-9 Entry
04/28/2025
Pre-Fire Plan #
205
Fire Zone 18.11-0, 686' & 702' Elevation River Screen
House
Pre-Fire Plan
- 138
Fire Zone 11.4A-2, Auxiliary Building 383-0 Elevation Unit 2
2B Auxiliary Diesel Feedwater Pump and Day Tank Room
Pre-Fire Plan
- 198
Fire Zone 18.3-2 Miscellaneous Area 377'-0" Elevation Unit
2, Main Steam and Auxiliary Feedwater Pipe Tunnel
Pre-Fire Plan #74
Fire Zone 8.5-2 Northeast, Turbine Building 426 Elevation
Unit 2 General Area - NE
Pre-Fire Plan #75
Fire Zone 8.5-2 Northwest, Turbine Building 426 Elevation
Unit 2 General Area - NW
Pre-Fire Plan #76
Fire Zone 8.5-2 Southeast, Turbine Building 426 Elevation
Unit 2 General Area - SE
Pre-Fire Plan #77
Fire Zone 8.5-2 Southwest, Turbine Building 426 Elevation
Unit 2 General Area - SW
Fire Plans
Pre-Fire Plan #96
Fire Zone 11.1A-0 Auxiliary Building 330'-0" Elevation Unit 1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Essential Service Water Pump Room
Pre-Fire Plan #97
Fire Zone 11.1B-0 Auxiliary Building 330'-0" Elevation Unit 2
Essential Service Water Pump Room
2B AF Pp HXs SX Rtrn VLV Stroke Issue
10/24/2023
Corrosion on AB 383 Floor Plug 0DSP372 Lifting Shackles
03/24/2025
Corrective Action
Documents
Rigging Evaluation AB 383 Fl Plug 0DSP372 Lifting
Shackles
03/25/2025
Drawings
M-126, Sheet 1
Diagram of Essential Service Water
BH
Engineering
Changes
Temporary Change to Remove Disc from Inside of Gate
Valve 2SX175
000
BOP SX-22
Essential Serviced Water Leak Isolation
Procedures
OP-BY-102-106
Operator Response Time Program at Byron Station
U2 Train A AF Valves Position Indication Test
01/25/2025
Work Orders
U2 Train B AF Valves Position Indication Test
2/02/2025
Boric Acid Identified on LSIV 2RC8001C
10/09/2023
B2R24 - Relevant Condition on 2FW09002C Pipe Support
10/12/2023
Through Wall Leak Upstream of 2SX035
11/11/2023
Trend - Missing Documentation for ASME R/R Work
Packages
05/14/2024
ASME Replacement WO 5322470 Documentation
Incomplete
06/05/2024
Boric Acid Pump Berm Filling with Crystals
07/30/2024
NRC ID: N-722-1 Augmented ISI Program Implementation
Issues
09/17/2024
Inactive Boric Acid Leak at Bolted Connection of 2RH8703A
10/03/2024
Repeat Boric Acid Leakage - 2CV206 Leak-by Past Valve
Seat
11/07/2024
Active Borated Water Leak - Fitting Downstream of
2PS9352A
11/19/2024
Active Leak on 2CS049A
2/20/2025
B2R25 ISI Exams - Relevant Condition of 2SI01009X
Support
04/17/2025
Corrective Action
Documents
B2R25 ISI Exams - Relevant Conditions Identified on
04/17/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2SI23009X Support
Engineering
Changes
Op Eval 23-001 - Pinhole Leak on Tee Branch Weld - Line
2SX32A-4" Use of Code Case N-513 (IR 04716984)
11/15/2023
Bare Metal Visual Examination of RPV Closure Head
10/16/2023
Magnetic Particle Examination of 2RC-01-R/RVHC-01
Reactor Vessel Head to Flange Weld
04/20/2025
Ultrasonic Examination of 2RC-28A-3/J03
04/18/2025
Ultrasonic Examination of 2RC-28A-3/J04
04/18/2025
Ultrasonic Examination of 2RC-28A-3/J05
04/18/2025
Ultrasonic Examination of 2RC8001A LSIV Studs
04/21/2025
Ultrasonic Examination of 2RC-01-R/RVHC-02 Meridian
04/21/2025
Bare Metal Visual Examination of Reactor Vessel Bottom
Head and Instrument Penetrations
04/16/2025
Bare Metal Visual of Reactor Pressure Vessel Closure Head
04/22/2025
VT-1 Examination of 2AF-01-PA/E-1 Welded Attachment
04/17/2025
VT-3 Examination of 2SI08FA-3/2SI23009X Piping Support
04/18/2025
NDE Reports
VT-3 Examination of 2SI05DA-6/2SI01009X Piping Support
04/18/2025
Magnetic Particle (MT) Examination
VT-1 Visual Examination
VT-3 Visual Examination of Component Supports,
Attachments, and Interiors of Reactor Vessels
Ultrasonic Examination of Ferritic Piping Welds
Ultrasonic Examination of Austenitic Piping Welds
Manual Ultrasonic Examination of Reactor Pressure Vessel
Ultrasonic Examination of Studs and Bolts
Bare Metal Visual Examination for Nickel Alloy Materials
WDI-SSP-1265
Ultrasonic Examination of Loop Stop Isolation Valve Studs at
Byron and Braidwood
WPS 1-1-GTSM-
ASME Welding Procedure Specification
Procedures
WPS 8-8-GTSM
ASME Welding Procedure Specification
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Through Wall Leak Upstream of 2SX035
08/16/2024
Work Orders
LR-Support Eddy Current Testing for 2A CS Pump Cubicle
Cooler
01/30/2025
OP-A-101-111-
1001
Operations Standards and Expectations
Procedures
Conduct of Simulator Training and Evaluation
Conditional Release of Switch CID 1718662-1
2/27/2024
Conditional Release for Cat ID 1423365-2
05/08/2024
Conditional Release for Cat ID 44508-2
07/24/2024
Conditional Release of Cable for FRV EC 638421
09/13/2024
Conditional Release for Cat ID 1838928-4
09/15/2024
Conditional Release for Cat ID 0001428646-2, Cooler
09/21/2024
Conditional Release for Cat ID 668536-2
09/22/2024
1LT-0548 (1D SG Level Narrow Range) 1PA03J Card
Failure
2/20/2024
0A SX Makeup Pump Upper Fuel Sightglass Empty
01/21/2025
NRC Identified Injector Reinstall On 1B AF Pump
Undocumented
03/22/2023
Framatome Notification of Part 21 limitorque Torque Switch
03/24/2025
NOS Finding: Unqualified Grease Available and
Possibly Used
03/28/2025
Issue Report Incorrectly Screened for MRule Applicability
04/01/2025
Issued Grout from Stores That Was Expired
04/01/2025
0A SX M/U Pump Failed to Start
04/24/2025
QV Identified Damaged Flex Hose
04/26/2025
Corrective Action
Documents
Conditional Release for Cat ID 1843305-1 A2 Card
04/27/2025
Corrective Action
Documents
Resulting from
Inspection
NRC IR - Follow Up IR Not Timely Generated
05/30/2025
Miscellaneous
2-190147
AMETEK Notification of Potential Defect - 10 CFR Part 21
for Charger Control Board X302, 80-9214031-90
A
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
DT258613
GE Over/Under Voltage Time Delay Relays
04/08/2025
NPC-2025-009
Framatome Notification of Part 21 (Limitorque Torque Switch
P/N 11501-010, FRA P/N 1167723-662)
03/05/2025
BMP 3222-2
Preventive Maintenance of Essential Service Water
Make-Up Pump Drive Unit
Maintenance Rule Implementation Per NEI 18-10
Maintenance Rule 18-10 - Failure Definition
Maintenance Rule (a)(1) and (a)(2) Requirements
Maintenance Rule 18-10 - Expert Panel Roles and
Responsibilities
Quality Material/Components Control and
Identification/Segregation of Non-Conforming Items
Work Execution & Close Out
Compression Fittings Inspection, Installation, Remake and
Repair
Warehouse Operations
Procurement Engineering Support Activities
Procurement Engineering Process and Responsibilities
SM-AA-300-
1001-F-03
Procurement Engineering and Design Engineering Interface
Agreement
Procedures
WC-AA-120-F-01
Work Orders
EWP-FNM 0A SX Makeup Pump Fuel Priming
01/23/2025
Unexpected Alarm DC 112 Grounds
06/19/2025
Unexpected Alarm - DC 112 Ground
06/25/2025
NRC Identified - FLEX Pump Staged Without Hoses
04/16/2024
2A MSIV EH Level Approaching Rounds Minimum Level
05/05/2025
1A MSIV EH Level Approaching Minimum
05/05/2025
2A MSIV EH Level Trending Down
05/08/2025
2A MSIV Hydraulic Oil Level High
05/10/2025
Unexpected Alarm Received in MCR (1-1-C5)
05/10/2025
1B MSIV Alarm Received in MRC (1-1-C5)
05/22/2025
Corrective Action
Documents
05/26/2025
Procedures
CC-BY-118-1004
Byron Station Units 1 & 2 Final Integrated Plan Document
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Mitigation Strategies for a Beyond-Design-Basis External
Event
Mild Corrosion on 0SX163F Nuts and Bolts
04/10/2023
Moderate Corrosion on 0SX163G
04/21/2023
Sets of Nuts and Bolts Not Replaced on 0SX163F
05/06/2024
Replace Sets of Nuts & Bolts Connected to VLV Body on
0SX163F
08/14/2024
1SX016B Diagnostic Testing Results
09/27/2024
1SX027B Diagnostic Testing Results
09/27/2024
1(2)SX016A/B & 1(2)SX027A/B Inconsistent Testing
Conditions
09/27/2024
1AF005 Failed Stroke Time Test
03/07/2025
1AF005E Open Stroke Time Less Than IST Limit
03/20/2025
Actuator Rebuild Needed
03/21/2025
OOT on Temporary Transmitter
04/19/2025
OOT on Temporary Transmitter 2FE-0925
04/18/2025
DC 211 Battery Voltage Low (Cell 44)
05/28/2025
Corrective Action
Documents
Moderate to Heavy Corrosion on 0SX163G Flange Studs
and Nuts
06/09/2025
Corrective Action
Documents
Resulting from
Inspection
NRC ID: Discrepancies on Capacity Test Docs: Battery Bank
211
06/18/2025
Evaluation of Removal and Replacement of One Stud
at a time For The 0SX163 Valve Flanges
Evaluation for Removal of Up to Three Adjacent Studs
on 0SX163A-H
Engineering
Changes
TCC Temporarily Jumper out Cell 44 of ESF Battery 211 to
Allow 125 VDC ESF Battery 211 Operation with 57 Cells
Miscellaneous25-001
IST Valve Evaluation Form 1AF005E
03/20/2025
1BOSR 0.5-
2.AF.1-2
Unit One Train B Auxiliary Feedwater Valves Stroke Test
Procedures
2BOSR
5.5.8.CV.5-1C
Unit Two Comprehensive Inservice Testing (IST)
Requirements for Centrifugal Charging Pump 2CV01PA
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2BOSR
5.5.8.CV.5-2C
Unit Two Comprehensive Inservice Testing (IST)
Requirements for Centrifugal Charging Pump 2CV01PB
2BOSR
5.5.8.SI.2-1
Unit Two IST Check Valve Stroke Test for 2SI8819A/B/C/D
2BOSR 5.c.2-1
Unit Two Charging/Safety Injection System Flow Balance
Diagnostic Testing of Motor Operated Valves
PM Inspection, In-Service Diagnostic Test, PIT
01/01/2025
PM Inspection, In-Service Diagnostic Test, PIT
2/07/2025
EWP EM 125V Battery Bank 5Yr Capacity Test 2DC01E
04/22/2025
2CV01PB Comprehensive IST Requirements for Charging
Pump
04/23/2025
Work Orders
03/20/2025
NRC Identified - FLEX Pump Staged Without Hoses
04/16/2025
Corrective Action
Documents
2B SG PORV Lifted During Plant Power Ascension
05/01/2025
Miscellaneous
B2R25 Shutdown Safety Plan
2BGP 100-3
Power Ascension
110
2BGP 100-3T1
Power Ascension Flowchart
CC-BY-118-1004
Byron Station Units 1&2 Final Integrated Plan Document
Mitigation Strategies for a Beyond-Design-Basis External
Event (NRC Order EA-12-049)
Risk Management Documentation
On-Line Risk Management
Procedure Use and Adherence
Protected Equipment Program
Shutdown Safety Management Program
Shutdown Safety Management Program Byron / Braidwood
Annex
Procedures
On-Line Work Control Process
Work Orders
Byron Unit 2 PreOutage Tasks
2AF005B Stuck Midposition
04/07/2025
Gulled Motor Shaft on 02-SX01PA
04/18/2025
Local Leak Rate on Pen-39 at Warning Limit
04/20/2025
Corrective Action
Documents
2A SX Motor WO Missing Info to Wire Up the New Motor
04/21/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2A DG Unable to Reach Overspeed Trip Setpoint
04/21/2025
2A DG Unable to Reach Overspeed Trip Setpoint (04/22/25)
04/22/2025
Emergent Scaffold for LLRT Access to 2PR061
04/23/2025
2RE9170 Limit Switch Issue
04/23/2025
2BOSR 8.1.11-1 UV Relay Resistance
04/24/2025
2RF027 Valve Failed Open with the MCR Switch Demanding
Closed
04/27/2025
TD-1 Not Within Limits SSPS Testing
04/28/2025
1B DG Upper Lube Oil Cooler Leak
05/07/2025
05/07/2025
1B DG Starting Air Receiver Pressure Uneven Change
on Start
05/07/2025
1A DG Damaged Push Rod
05/19/2025
Damaged Grease Fitting 1DG5048A
05/19/2025
1A DG Supply Fan Torn Expansion Boot
05/19/2025
1A DG Cylinder 5L Power Head Appears to be Cracked
05/19/2025
1DG01KA-X1 Upper Cooler
05/19/2025
1DG01KA-X2 Lower Cooler Coating
05/19/2025
1DG01KA OOT
05/19/2025
Trend IR - Restricted Tube in 1DG01KA-X1
05/21/2025
Damaged Ferrel on Fuel Line for 1A DG
05/22/2025
1A DOST Substrate Material Condition
05/22/2025
1SX169A Breaker will not Close
05/24/2025
Request Permission to Turn Down the SX Motor Mounting
Bolts to Allow Ease of Alignment
04/23/2025
Engineering
Changes
Request Permission to Reduce/Machine the SX Motor
Coupling Spacer to Allow for Proper Float
04/22/2025
DA27372
Safety Valve Test Data
04/04/2025
Miscellaneous
DA27374
Safety Valve Test Data
03/13/2025
0BOSR
5.5.8.SX.5-2c
Unit Zero Comprehensive Inservice Testing (IST)
Requirements for Essential Service Water Makeup Pump 0B
BOSR 8.1.2-2
Unit One 1B Diesel Generator Operability Surveillance
Procedures
2BOSR 0.5-
Unit Two Auxiliary Feedwater Valves Train A Indication Test
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2.AF.3-1
2BOSR 3.1.5-1
Unit Two Train A Solid State Protection System Surveillance
2BOSR 5.2.5-1
ECCS Subsystem Automatic Valve Actuation Test
2BOSR
5.5.8.SX.5-1C
Unit Two Comprehensive Inservice Testing IST
Requirements for Essential Service Water (SX) Pump
2SX01PA and Unit 2 SX Pumps Discharge Check Valves
2BOSR 6.1.1-10
Unit Two Primary Containment Type C Local Leakage Rate
Tests and IST Tests of Reactor Building Drains and Vents
System;
2BOSR 6.1.1-12
Unit Two Primary Containment Type C Local Leakage Rate
Tests and IST Tests of Component Cooling System
2BOSR 6.1.1-15
Unit Two Primary Containment Type C Local Leakage Rate
Tests and IST Tests of the Instrument Air System
2BOSR 6.1.1-20
Unit Two Primary Containment Type C Local Leakage Rate
Tests and IST Tests of Process Radiation Monitoring System
2BOSR 6.1.1-22
Primary Containment Type C Local Leakage Rate Tests
and IST Test of Safety Injection System
2BOSR 8.1.11-1
Unit 2 2A Diesel Generator Sequencer Test
2BOSR DG-R1a
Unit Two 2A Diesel Generator Overspeed Trip
Surveillance - 18 Month
2BOSR XII-1
Gaseous Leak Testing of the 2RH01SA/B Valve
Containment Assemblies
2BSRV 7.1.1-1
Unit 2 Main Steam Safety Valves Operability Test
BOP DG-11T2
Diesel Generator Operating Log
BOP RH-5
RH System Startup for Recirculation
NF-BY-510
Low power Physics Test Program
LLRT for P-65 - 2RE9160A, 2RE9160B, and 2RE9157
04/25/2025
2MS017B IST Trevitest
04/10/2025
Replace Governor Actuator
04/23/2025
Replace Governor Digital Reference Unit
04/21/2025
Rebuild Actuator and Replace Regulator
04/29/2025
2301A Electronic Governor
04/23/2025
Work Orders
LLRT for P-52 - 2PR001A and 2PR001B
04/28/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
LLRT for P-11 - 2RE1003, 2RE022, and 2RE92170
04/24/2025
2A Diesel Generator Sequencer Test
04/24/2025
STT/PIT for 2IA065 and 2IA066
04/29/2025
Train A SSPS Surveillance
04/29/2025
Perform 15 YR PM and Overhaul
04/27/2025
Low Power Physics Testing Program Using the W ADRC
05/01/2025
As Found LLRT for P-21 - 2CC9534, 2CC9416, and
2CC9414
04/26/2025
As Found LLRT for P-39 - 2IA066, 2IA065, and 2IA091
04/29/2025
CM-2MS014B-Remove/Install 2B S/G MS Outlet HDR
ATMOS RLF VLV
2/05/2025
CM - 2MS017 - Remove/Install 2B S/G MS OUTLET HDR
ATMOS RLF VLV
04/20/2025
2MS013C IST Trevitest
04/10/2025
2MS013D IST Trevitest
04/10/2025
2MS014C IST Trevitest
04/10/2025
2MS015A IST Trevitest
04/10/2025
2MS015B IST Trevitest
04/10/2025
2MS015C IST Trevitest
04/10/2025
2MS017C IST Trevitest
04/10/2025
BOSR 8.1.2-2, 1B D/G Semi-Annual Run
05/07/2025
0SX02PB Comprehensive IST REQ for SX Makeup Pump
05/30/2025
U2 Train A AF Valves Position Indication Test
04/07/2025
2RE1003 Will Not Go Open
04/02/2025
CM 2MS016B - Remove/Install U2 S/G MS OUTLET HDR
ATMOS RLF VLV
04/24/2025
Corrective Action
Documents
Byron 2025 Semi-Annual Health Physics Drill
05/14/2025
Procedures
Drills and Exercise Program
BY-2-25-00613
Rx Head Dis-Assembly / Re-Assembly
Original
BY-2-25-00614
RX Head / Upper Internals Moves
Original
BY-2-25-00623
RX Cavity Decon
Original
ALARA Plans
BY-2-25-00654
Seal Table - All Activities
Original
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Emergent Scaffold & Lead Needed to Reduce Background
Argos
04/15/2025
U2 CAF Northwest Roll-Up Door (0DSSD301) Stuck Halfway
Open
04/15/2025
Elevated Dose Rates on Pipe in VCT Tank Room
04/15/2025
CAM Alarm Following Release of CNMT
04/16/2025
Active Leak on the 2PS166
04/17/2025
Corrective Action
Documents
B2R25: Potential Trend in Rad Worker Behaviors
04/19/2025
Miscellaneous
Alpha Area Level Assessment
04/18/2024
2BGP 100-6T5
Changing Setpoints of 2RT-AR011 and 2RT-AR012
Operational ALARA Planning and Controls
Radiological Risk Management
Controls for the Draining and Decon of BWR/PWR Reactor
Cavity and Associated Pits
RP-BY-300-1023
Radiological Controls for Reactor Head and Upper Internals
Movement
Procedures
RP-BY-905
1(2)RE-AR011(12) Fuel Handling Incident Monitor Setpoint
Change
16-Apr-25-
210049
Reactor Head Lift 426 Elevation Air Sample Gamma
Spectrum Analysis
04/16/2025
25-238729
Conoseals Inital/Downpost Survey
04/14/2025
25-238732
Seal Table Room Downpost to Radiation Area
04/14/2025
25-238870
Unit 2 Reactor Cavity Sandbox Cover Removal
04/25/2025
25-239034
1PS34T/2PS34T Drain Tanks
04/16/2025
25-239146
Reactor Head Stand Area Before and After Shielding
Installation
04/17/2025
2-Apr-25-
210012
Unit 2 CAF Air Sample Gamma Spectrum Analysis
04/22/2025
Radiation
Surveys
NISP-RP-006,
Personnel Contamination Event Log (PCE #2)
04/22/2025
BY-2-25-00613
RX Head Dis-Assembly / Re-Assembly
Radiation Work
Permits (RWPs)
BY-2-25-00614
RX Head / Upper Internals Move
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
BY-2-25-00623
RX Cavity Decon
BY-2-25-00654
Seal Table: All Activities
Corrective Action
Documents
Resulting from
Inspection
NRC ID: Gap in MSPI Recorded Data
06/11/2025
NRC Performance Indicator Data; Mitigating Systems -
Safety System Functional Failures
April 2024
through
March 2025
NRC Performance Indicator Data; Mitigating Systems -
Mitigating System Performance Index for Emergency AC
Power
April 2024
through
March 2025
Miscellaneous
NRC Performance Indicator Data; Mitigating Systems -
Mitigating System Performance Index for High Pressure
Injection
April 2024
through
March 2025
71151
Procedures
Collecting and Reporting of NRC ROP Performance
Indicator Data
Byron IST 5th 10YR Interval Update
01/26/2024
Time Requirement Guidelines for Performing UTs
2/07/2025
Corrective Action
Documents
NRC Identified Incorrect Procedure Revision Used
04/10/2025
Engineering
Changes
MSSV Testing Thermal Equilibrium Acceptance Eval
04/16/2025
BMP 3114-21
Main Steam Safety Valve Testing Using Setpoint Verification
Device
BMP 3114-21
Main Steam Safety Valve Testing Using Setpoint Verification
Device
BVP 800-30
Essential Service Water Fouling Monitoring Program
(GL 89-13 Program Basis Document)
Procedures
Procedure Use and Adherence
NRC Identified Discrepancy in Protected Equipment Posting
NRC Identified Protected Equipment Not Posted Correctly
B1R26 LL - Protected Equipment Discrepancy Identified
NRC Identified - Protected Equipment Discrepancies
04/16/2025
Corrective Action
Documents
Lesson Learned for B2R25 Protected Equipment
04/16/2025