IR 05000456/2025010

Comprehensive Engineering Team Inspection (Ceti) Report 05000456/2025010 and 05000457/2025010

ML25239A583
Person / Time
Site:	Braidwood
Issue date:	09/04/2025
From:	Jasmine Gilliam NRC/RGN-III/DORS/EB1
To:	Rhoades D Constellation Energy Generation
References
IR 2025010
Download: ML25239A583 (1)

Text

SUBJECT:

BRAIDWOOD STATION - COMPREHENSIVE ENGINEERING TEAM INSPECTION (CETI) REPORT 05000456/2025010 AND 05000457/2025010

Dear David Rhoades:

On July 28, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Braidwood Station and discussed the results of this inspection with Adam Schuerman and other members of your staff. The results of this inspection are documented in the enclosed report.

Six findings of very low safety significance (Green) are documented in this report. Six 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 Braidwood 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 Braidwood Station.

September 4, 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, Jasmine A. Gilliam, Chief Engineering Branch 2 Division of Operating Reactor Safety Docket Nos. 05000456 and 05000457 License Nos. NPF-72 and NPF-77 Enclosure:

As stated cc w/ encl: Distribution via LISTSERV Signed by Gilliam, Jasmine on 09/04/25

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a Comprehensive Engineering Team Inspection (CETI) at Braidwood 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 Evaluate the Degraded Voltage Condition for the Class 1E Battery Charger Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-01 Open/Closed None (NPP)71111.21M The inspectors identified a finding of very low safety significance (Green) and a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B,

Criterion III, Design Control, for the licensees failure to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures and instructions. Specifically, in Calculation 19-AQ-68, the licensee failed to evaluate the capability of Battery Charger 112 to meet its design requirements during a design bases degraded voltage condition when the calculated input voltage results fell outside the assumed voltage range.

Failure to Implement and Maintain Beyond-Design Basis External Events Mitigation Strategies and Required FLEX Equipment in accordance with 10 CFR 50.155 Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-02 Open/Closed

[H.7] -

Documentation 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.155, Mitigation of Beyond-Design-Basis Events, for the licensees failure to implement and maintain mitigation strategies that maintain or restore core cooling, containment, and spent fuel pool (SFP) cooling capabilities.

Specifically, the inspectors identified the licensee failed to: (1) validate and document FLEX strategies were capable of being implemented site-wide and remained viable on a recurring basis, (2) maintain availability of the minimum required number of low and medium head FLEX pumps, (3) establish flowrate acceptance criteria for the low head FLEX pumps to ensure the pumps had sufficient capacity and capability to perform their required function.

Failure to Translate the Maximum Component Cooling Water Temperature into Transfer to Cold Leg Recirculation Emergency Operating Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-03 Open/Closed

[P.2] -

Evaluation 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to update Emergency Operating Procedures (EOP)1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, to reflect the new maximum component cooling (CC) water temperature following a design-basis loss-of-coolant-accident (LOCA). The new maximum CC temperature was a result of a License Amendment Request (LAR), approved on July 26, 2016, which increased the maximum ultimate heat sink (UHS)temperature.

Failure to Verify the Adequacy of the Essential Service Water System Design Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-04 Open/Closed

[H.6] - Design Margins 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to verify or check the adequacy of the essential service water (SX)system design to ensure all safety-related loads could be adequately cooled during accident conditions. Specifically, the site failed to establish design control measures such as SX system flow balance test procedures, hydraulic calculations, or alternate methods to verify the design capability of the Unit 1 and 2 SX system.

Failure to Evaluate a Single Failure of an Essential Service Water Pump Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-05 Open/Closed

[H.14] -

Conservative Bias 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to translate a limiting single failure of the essential service water (SX)system under post loss-of-coolant-accident (LOCA) conditions into specifications, drawings, procedures, and instructions design documents. Specifically, the licensee failed to assure a single SX pump on the accident unit was capable of providing sufficient cooling to all required SX heat loads accident conditions.

Failure to Identify Local Operation of Motor Operated Valve (MOV) 1(2)CC9415, After a Loss of Electrical Power, was Within Design Bases Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-06 Open/Closed None (NPP)71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.55a(f)(4)(ii), Applicable IST Code: Successive code of record intervals, when the licensee failed to identify local operation to close MOV 1(2)CC9415, using the handwheel, was required to comply with the site's design bases. Specifically, following certain design bases accidents or transients combined with a single failure of the electrical power supply to 1(2)CC9415, operators would need to locally close 1(2)CC9415 to isolate the non-essential (service loop) loads of the CC system and ensure sufficient CC cooling flow remains available to the essential (safety-related) loads. Additionally, the licensee's Inservice Testing (IST) program failed to include the full stroke tests required for credited manual valves.

Additional Tracking Items

Type Issue Number Title Report Section Status URI 05000456,05000457/20 25010-07 10 CFR 50.59 Evaluation Needed to Determine if Breaker Position Change for 1(2) CC9415 required NRC Approval 71111.21M Open

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 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

===71111.21M - Comprehensive Engineering Team Inspection The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:

Structures, Systems, and Components (SSCs) (IP Section 03.01)===

For each component sample, the inspectors reviewed the licensing and design bases including:

(1) the Updated Final Safety Analysis Report (UFSAR);

(2) the Technical Specifications (TS); and

(3) the Technical Requirements Manual (TRM). The inspectors reviewed a sample of operating procedures (including normal, abnormal and emergency procedures), overall system/component health (including condition reports and operability evaluations, if any) and associated maintenance effectiveness (e.g., Maintenance Rule, procedures). The inspectors performed visual inspections of the accessible components to identify potential hazards and/or signs of degradation. Additional component specific design attributes reviewed by the inspectors are listed below.

(1) Ultimate Heat Sink (UHS)

1. Protection against external events:

a.

Flooding b.

Seismic

3. Review of reservoir capacity:

a.

Review of underwater structures and/or excavations b.

Sediment intrusion c.

Volumetric Testing and/or Calculations

4. Applicable Design Bases Calculations:

a.

UHS Temperature limit b.

Required minimum volume

(2) Unit 1: Essential Service Water (SX) Pump 1A (1SX01PA)

1. Protection against external events:

a.

Flooding, including sump pump b.

Seismic c.

High Energy Line Break (HELB)

2. Mechanical design calculations and considerations:

a.

Flow capacity & balance b.

Minimum flow c.

Runout flow d.

Required Net Positive Suction Head (NPSH) and vortexing e.

Hydraulic transients (water hammer)

3. Test/inspection procedures, acceptance criteria, and recent results:

a.

Pump comprehensive Inservice Testing (IST) surveillances b.

Flow balance/capacity tests c.

Pump quarterly IST surveillances

4. Electrical design calculations and considerations:

a.

Overcurrent relay setpoint b.

Starting and running voltage at offsite degraded voltage c.

Brake horsepower/motor service factor

(3) Unit 1: BUS 141 (4160 Vac - 1SX Pump's Power Supply)

1. Test/inspection procedures, acceptance criteria, and recent results:

a.

TS surveillance b.

Relay calibration

2. Electrical design calculations and considerations:

a.

Loading calculations b.

Voltage regulation c.

Coordination calculations d.

Overcurrent protection e.

Protective devices and trip set points

(4) Unit 1: Bus 111 (125 Vdc - 1SX Pumps Control Power)

1. Test/inspection procedures, acceptance criteria, and recent results:

a.

TS surveillance

2. Electrical design calculations and considerations:

a.

Short circuit calculations b.

Coordination calculations c.

Overcurrent protection d.

Protective devices and trip set points

(5) Unit 1: Discharge Strainer 1A (for SX Pump 1A) & Associated MCC 131X5 (480 Vac Power Supply)

1. Protection against external events:

a.

Flooding, including sump pump b.

Seismic c.

HELB

2. Mechanical design:

a.

Debris loading impact to flow capacity/balance b.

Debris loading impact to structural design c.

Instrument differential pressure setpoints d.

Mesh size vs downstream component openings e.

Manual backwash capability

4. Automatic backwash design:

a.

Voltage drop b.

Degraded voltage effects c.

Emergency power (EDG)d.

Automatic backwash protective relays e.

Control logic

5. MCC Electrical design calculations and considerations:

a.

Loading calculations b.

Short circuit calculations c.

MCC capacity d.

Degraded voltage e.

Overcurrent protection f.

Loss of voltage g.

Protective devices h.

DC Voltage to MCC breakers

(6) Unit 1: Component Cooling Water (CC) Heat Exchanger (HX)

1. Design calculations and considerations:

a.

Minimum cooling water flowrate b.

Maximum cooling water temperature c.

Maximum working fluid temperature d.

Tube plugging limit e.

Heat transfer capacity

2. Test/inspection procedures, acceptance criteria, and recent results:

a.

Flowrates b.

Inspection or thermal performance test c.

Eddy current

(7) Unit 1: Motor Operated Valve (MOV) CC HX Outlet Isolation Valve (1SX007)

& Associated MCC 131X1 (480Vac)

1. Mechanical design

a.

Weak link analysis b.

Required thrust (torque)c.

Closure/Opening time d.

Maximum allowed leakage e.

Maximum differential pressure

2. Test/inspection procedures, acceptance criteria, and recent results:

a.

IST b.

TS Required Surveillance

3. Motor power requirements:

a.

Voltage drop b.

Control logic c.

Control voltage drop d.

Thermal overload e.

Required minimum voltage f.

Degraded voltage effects g.

Motor thermal overload protection h.

Emergency power (EDG)

4. MCC Test/inspection procedures, acceptance criteria, and recent results:

a.

Relay calibration

5. MCC Electrical design calculations and considerations:

a.

Short circuit calculations b.

Degraded voltage c.

Overcurrent protection d.

Protective devices e.

DC Voltage to MCC breakers

(8) Unit 1: Bus 131X (480 Vac - Power for MCC 131X1)

1. Test/inspection procedures, acceptance criteria, and recent results:

a.

Relay calibration

2. Electrical design calculations and considerations:

a.

Loading calculations b.

Short circuit calculations c.

Coordination calculations d.

Degraded voltage protection e.

Protective devices and trip set points

(9) Unit 1: Low-Head, Diesel Driven FLEX Pumps (0FX03PA and 0FX03PB)

1. Protection against external events:

a.

Seismic

2. Mechanical design calculations and considerations:

a.

Flow capacity & balance b.

Minimum flow c.

Runout flow d.

Required submergence e.

Water supply availability f.

Engine sizing g.

Fuel oil volume consumption h.

Fuel oil available volume/level i.

Engine trip set points

3. Test/inspection procedures, acceptance criteria, and recent results:

a.

capacity tests

Modifications (IP Section 03.02) (4 Samples)

(1) Engineering Change (EC) 622930, DC System Short Circuit Calculations Update

(2) EC 634937, Min Wall Thickness Calc for Lines 2SX05CA-6 and 2SX04DA-6 for UT Examination

(3) EC 636627, N-513-4 Code Case Evaluation for Pin Hole Leak at Weld on Line 0SX03A-30

(4) EC 624189, Breaker Position Change for MOVs 1(2)CC9415 to Eliminate Spurious Closure 10 CFR 50.59 Evaluations/Screening (IP Section 03.03) (10 Samples)

(1) BRW-E-2021-001, "TRM Change 20-10 - ECCS 1(2)SI8801B and 1SI8802B

(2) BRW-E-2023-001, TRM Change 23-001 - 1SX001A and 2SX001A

(3) BRW-S-2022-80, Retention of Temporary Scaffold Erected Under WO 05133256-15 beyond 90 days

(4) BRW-S-2023-013, Revise TRM Appendix M, Technical Specification Bases B3.8.3, and UFSAR Appendix F to Align Byron and Braidwood Station Diesel Fuel Oil Testing Programs

(5) BRW-S-2023-047, Breaker Position Change for MOVs 1(2)CC9415 to Eliminate Spurious Closure

(6) BRW-S-2023-046, Issue BYR 125VDC System ETAP Calculation and Issue BRW 125VDC System ETAP Calculation

(7) BRW-S-2023-033, Bus voltage acceptance criteria for SR 3.8.9.1

(8) BRW-S-2024-036, BwOP VC-20 revision to CDP and incorporate BwOP VC-19

(9) BRW-S-2025-018, Retention of PBI 24148 for turbine building louver tarps to remain beyond ninety-days

(10) BRW-S-2024-015, Braidwood SG License Renewal Commitment Update for Appendix F of the UFSAR

Operating Experience Samples (IP Section 03.04) (2 Samples)

(1) NER NC-07-044-R, Essential Service Water Piping Degradation

(2) World Association of Nuclear Operators (WANO) Significant Operating Experience Reports (SOER) 2007-02, Intake Cooling Water Blockage

INSPECTION RESULTS

Failure to Evaluate the Degraded Voltage Condition for the Class 1E Battery Charger Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-01 Open/Closed None (NPP)71111.21M The inspectors identified a finding of very low safety significance (Green) and a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures and instructions. Specifically, in Calculation 19-AQ-68, the licensee failed to evaluate the capability of Battery Charger 112 to meet its design requirements during a design bases degraded voltage condition when the calculated input voltage results fell outside the assumed voltage range.

Description:

On or about June 5, 2025, during the inspectors review of the Calculation 19-AQ-68, Division Specific Degraded Voltage Analysis, the inspectors identified a discrepancy between the calculated degraded voltage values and assumed range of allowed AC input values for safety-related Battery Charger 112 (1DC04E). The Battery Charger provides a float charge to the train B 1E Battery 112 and provides the DC supply to the 125VDC bus, source to instrumentation and control power during normal and post-accident conditions. The identified discrepancy was discussed with the licensee during the first on-site week, in particular during a meeting with design engineering on June 13, 2025.

Offsite power is the preferred source to supply the 1E 4160V bus. By Technical Specification (TS) Surveillance Requirement (SR) 3.3.5.2, the bus may degrade to 3930 V, at that point degraded voltage relays start the Emergency Diesel Generator and transfer the 1E 4160 V bus to the on-site source. The licensee developed Calculation 19-AQ-68 with the purpose to assure all components will perform in the degraded voltage condition while connected to the off-site power supply. The Electrical Transient Analyzer Program (ETAP) calculation analyzed the expected voltage on the connected transformers, 480V buses, motor control centers and component terminals after expected line losses. The nominal input AC voltage for the battery charger is 480V. The calculation assumed an input voltage range of 480V +/-10%, that is, a minimum voltage of 432V AC. However, the calculation determined the voltage at the input terminals of the Battery Charger 112 to be 429.4VAC when the 4160V bus is degraded to the worst case of 3930VAC. This gave reasonable doubt that the battery charger could support the required TS minimum output voltage of 120.1 VDC and the associated required DC load in the degraded voltage condition.

In accordance with the Updated Final Safety Analysis Report (UFSAR) Section 3.1.2.2.8, Evaluation Against Criterion 17 - Electric Power Systems, the licensee adheres to IEEE Standard 308-1974 to meet design requirements for the engineered safety features of the electric power systems. In addition, UFSAR Section 8.3.2.1.1, Class 1E 125VDC Power System, and Standard IEEE 308-1974 establish that each battery charger is capable of floating the battery on the bus or recharging a completely discharged battery while supplying the largest combined demands of the various steady-state loads under all plant operating conditions. Standard IEEE 308-1974 also states, in part, Each battery charger supply shall furnish electric energy for the steady-state operation of connected loads required during normal and post-accident operation while its battery is returned to or maintained in a fully charged state.

Based on the above, the inspectors concluded the licensee failed to evaluate the degraded voltage condition for the Class 1E Battery Charger 112 in Calculation 19-AQ-68, Division Specific Degraded Voltage

Analysis.

Corrective Actions: On June 19, 2025, the licensee entered these issues into their corrective action program and assigned actions to their design engineering department to establish a basis for the operation of the charger at current voltage or reperform an analysis as necessary and document the results in a revision to Calculation 19-AQ-68. The licensee contacted the vendor, who provided Qualification Report, CL-QR-15346. It included performance tests for the Braidwood chargers, which determined a minimum input voltage of 414 VAC would generate the required output DC voltage and current. Therefore, the postulated input voltage of 429.4 VAC during a degraded condition would meet the chargers requirements.

Corrective Action References: AR 048747048747 NRC ID CETI Charger 112 potential low Voltage

Performance Assessment:

Performance Deficiency: The licensee failed to evaluate the design bases degraded voltage condition for the Class 1E Battery Charger 112. This was contrary to Standard IEEE 308-1974, UFSAR Section 8.3.2.1.1, Title 10 CFR 50, Appendix B, Criterion III and a performance deficiency. Specifically, Calculation 19-AQ-68, Division Specific Degraded Voltage Analysis, did not evaluate the performance of Battery Charger 112 when the calculated input voltage, during a postulated degraded voltage condition, fell below the assumed 10 percent input voltage range.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control 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 failure to evaluate the performance of the Battery Charger 112 in a design basis degraded voltage condition adversely affected the cornerstone objective of ensuring the capability of the charger to supply the battery and 125 VDC system to respond to initiating events to prevent undesirable consequences. In addition, this issue was comparable to the more than minor criteria described in Example 3.a of IMC 0612, Examples of Minor Issues.

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 screened the finding to Green (very low safety significance) because the finding did not result in the loss of operability or PRA functionality. This was based on answering Yes to question A.1 of the Exhibit 2 - Mitigating Systems Screening Questions.

Specifically, the licensee concluded the charger remained operable based on the vendor-provided Qualification Report, CL-QR-15346, which determined the battery charger would be capable of generating the required output DC voltage and current under the calculated degraded input voltage.

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

Enforcement:

Violation: Title 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures and instructions.

UFSAR Section 8.3.2.1.1, Class 1E 125VDC Power System," states in part, that each battery charger is capable of floating the battery on the bus or recharging a completely discharged battery while supplying the largest combined demands of the various steady-state loads under all plant operating conditions.

T.S. SR 3.3.5.2.b, establishes a degraded voltage allowable value of equal or greater than 3,930 Volts AC.

Design Calculation 19-AQ-68, Division Specific Degraded Voltage Analysis, Revision 7, analyzed the terminal voltages during a degraded voltage condition, including the input voltage for the safety-related Battery Charger 112.

Contrary to the above, as of June 26, 2025, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures and instructions. Specifically, in calculation 19-AQ-68, the licensee failed to evaluate the capability of Battery Charger 112 to meet its design requirements during a design bases degraded voltage condition when the calculated input voltage results fell outside the assumed voltage range.

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

Failure to Implement and Maintain Beyond-Design Basis External Events Mitigation Strategies and Required FLEX Equipment in accordance with 10 CFR 50.155 Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-02 Open/Closed

[H.7] -

Documentation 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.155, Mitigation of Beyond-Design-Basis Events, for the licensees failure to implement and maintain mitigation strategies that maintain or restore core cooling, containment, and spent fuel pool (SFP)cooling capabilities. Specifically, the inspectors identified the licensee failed to:

(1) validate and document FLEX strategies were capable of being implemented site-wide and remained viable on a recurring basis,

(2) maintain availability of the minimum required number of low and medium head FLEX pumps,

(3) establish flowrate acceptance criteria for the low head FLEX pumps to ensure the pumps had sufficient capacity and capability to perform their required function.

Description:

Braidwood Units 1 and 2 developed a final integrated plan (FIP) for mitigation strategies of a beyond-design basis external event (BDBEE) to comply with NRC Order EA-12-049. This order was issued following the event at Japans Fukushima Daiichi Nuclear Power Plant.

Braidwood established their FIP strategies, typically referred to as FLEX strategies, in Procedure CC-BR-118-1004, Braidwood Station Unit 1 & 2 Final Integrated Plan Document, Revision 1. Procedure CC-BR-118-1001, Site Implementation of Diverse and Flexible Coping Strategies (FLEX) and Spent Fuel Pool Instrumentation Program, Revision 9, described the implementation of the sites FIP.

FLEX strategies were required to mitigate a simultaneous loss of all alternating current (AC)power and loss of normal access to the ultimate heat sink (UHS). The three phase approach initially relied on installed plant equipment (phase 1), then transitioned to onsite portable FLEX equipment (phase 2), and then obtained offsite resources (phase 3). The sites phase 2 strategy, in part, deployed a low head FLEX pump which took draft suction from the UHS and used flexible hoses routed to a medium head FLEX pump to supply water to the spent fuel pool (SFP) for makeup and/or to provide back-up steam generator makeup in the event the diesel driven auxiliary feed pump could no longer perform its function. The inspectors review focused on the low head FLEX pumps (0FX03PA and 0FX03PB) which were required in phase 2.

The required FLEX pumps and water supplies were described in Procedure CC-BR-118-1004, Section 2.3.10.1. The low head FLEX pumps were trailer-mounted, diesel-driven centrifugal pumps that were required to implement the SFP makeup and reactor core cooling strategy. One low head FLEX pump was required to provide a water supply of 300 gallons per minute (gpm) to each Unit and 500 gpm to the SFP makeup simultaneously, for a total required flow of 1,100 gpm. Two low head FLEX pumps were available onsite to ensure reliability and availability of the required FLEX equipment as described in CC-BR-118-1004, Section 2.18.6. Each pump was rated at 1500 gpm at 150 psig.

As a result of this inspection, the inspectors identified three separate deficiencies as described and dispositioned below.

1. Validation of FLEX strategies

Procedure CC-BR-118-1004, Section 2.17, stated, The FLEX related procedures were walked down for accuracy and will be walked down at least once every five years to ensure the ability to perform the strategy remains unchanged Snow and Ice plans, which adequately address FLEX deployment travel routes, have been validated and will be validated at least once every three years.

Procedure CC-BR-118-1004, Section 2.18.1, described the programmatic elements of the sites FLEX strategy and stated, in part, Braidwood Station procedure CC-BW-118-1001 provides a description of the Site Implementation of Diverse and Flexible Coping Strategies (FLEX) and Spent Fuel Pool Instrumentation Program The key elements of this procedure include: Validation of FLEX Strategies. Procedure CC-BR-1001, Section 4.1.3, Validation of FLEX Strategies, stated, in part:

B. The FLEX related procedures shall be walked down for accuracy at least once every eight years to ENSURE the ability to perform the strategy given the current configuration of the plant, that component labels are in place and legible, and the referenced procedures credited in the site specific FIP remain viable to support the strategy.

C. Verify snow plans adequately address FLEX deployment travel routes at least once every four years.

The inspectors noted the discrepancies between the frequency of the two procedures stated the validation walkdowns would be performed. The site initiated corrective action document (IR) 04877265 for these discrepancies and planned to clarify and revise the strategy. The inspectors requested the last documented FLEX validations and questioned the actual frequency of walkdowns. In response, the site failed to locate any documented FLEX validations, notified the NRC inspection team, and on June 26, 2025, initiated IR 04876866.

This IR documented the sites failure to perform and document the FLEX validation walkdowns as stated in Procedures CC-BR-118-1004 and CC-BR-118-1004. The site planned to clarify the frequency and incorporate the required walkdowns into their process.

The inspectors determined this was a failure to implement and maintain FLEX mitigation strategies as required by 10 CFR 50.155(b)(1)(i).

2. Unavailability of low and medium head FLEX pumps

The inspectors reviewed CC-BR-118-1001, Attachment 9, Braidwood Flex Fuel Consumption Study, Revision 1, and noted the low head FLEX pumps were credited with a 15.6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> run time per engine before diesel fuel oil refueling was required. The conclusions of this study stated, in part, major FLEX equipment would not need refueling for approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from when the equipment was originally started. This ensured equipment would have sufficient fuel for the mitigation strategy timeline of a BDBEE. For this reason, FLEX pump engines were required to be maintained above a 90 percent fuel oil level. This acceptance criteria was included in surveillance procedures post run equipment storage requirements. Specifically, the inspectors reviewed Procedure 0BwOS FX-12, FLEX Pump Full Flow Test Surveillance, Revision 3, Section G. 2. Post run equipment storage condition acceptance criteria.

On June 12, 2025, the inspectors and licensee performed a walkdown of the FLEX storage building. The inspectors identified both low head FLEX pump engine fuel oil levels were below the required 90 percent level. This failed to meet the minimum requirement of one low head FLEX pump. As a result, the site entered the administrative action requirement (AAR)which tracked the unavailability of FLEX equipment required to support FLEX strategies.

Specifically, Procedure 0BwOS FX-1a, AAR FLEX Equipment, Revision 6, Attachment 2, Low Head FLEX Pump (0FX03PA or 0FX03PB), Condition E was entered. Condition E required a corrective action document to be initiated to determine the cause of the loss of function protection AND the initiation of actions to restore functionality or implement compensatory measures within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The site initiated IR 04872639 to document this issue and refilled the low head FLEX pump engines to above 90 percent within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Condition E required actions were then completed and the AAR was exited.

Additionally, during the extent of condition, the site identified three of three medium head pumps (0A, B, C) and one of three high head pumps (0C) were also below the 90 percent fuel oil level. Per the FIP, two medium and high head pumps were required to be available. As a result, the associated AAR Conditions were entered and the engines were refilled.

The inspectors determined this was a failure to ensure equipment relied upon to perform FLEX mitigation strategies had sufficient capacity and capability to perform their required functions as required by 10 CFR 50.155(c)(1).

3. Flowrate acceptance criteria for low head FLEX pumps

The sites FIP established the required flowrate of the low head FLEX pumps to be 300 gpm to each Unit and 500 gpm to the SFP makeup simultaneously, for a total required flow of 1,100 gpm. The inspectors reviewed Procedure 0BwOS FX-12, FLEX Pump Full Flow Test Surveillance, Revision 3, Section G. Acceptance Criteria, and noted the low head FLEX pump minimum required flow was greater than or equal to 900 gpm. On June 26, 2025, the inspectors questioned why this acceptance criteria was less than 1,100 gpm credited in the FIP.

Procedure 0BwOS FX-12, Section E. Limitations and Actions, noted the minimum low head FLEX pump flowrate of greater than or equal to 900 gpm was established per Calculation BRW-14-0030-M, Revision 0, Attachment B. The inspectors reviewed Calculation BRW-14-0030-M, Godwin Pump Suction Line Hydraulic Analysis to Support FLEX, Revision 1. The method and analysis section stated, The FIP states that the low head FLEX pump is sized to provide a water supply of 300 gpm to each Units Steam Generators and 500 gpm to SFP makeup simultaneously for a total of 1,100 gpm Based on those requirements, 1,100 gpm is a bounding flow rate. As a result of the inspectors questions, on June 26, 2025, the site initiated IR 04876831 which noted the previous calculation, Revision 0, had a lower acceptable flow value of 750 gpm. The site planned to initiate actions to revise the acceptance criteria based on the current revision of BRW-14-0030 and CC-BR-118-1004.

The inspectors determined this was a failure to ensure equipment relied upon to perform FLEX mitigation strategies had sufficient capacity and capability to perform their required functions as required by 10 CFR 50.155(c)(1).

Corrective Actions: The site generated three corrective action documents that addressed each example of the failure to implement and maintain BDBEE mitigation strategies and ensure equipment relied on for the mitigation strategies had sufficient capacity and capability to perform their required functions. These corrective actions, in part, included:

2. Entering the appropriate AARs and refueling all FLEX pump engines to above

90 percent fuel oil.

3. Planning to initiate actions to revise the low head FLEX pump flowrate acceptance

criteria.

Corrective Action References: 1. IR 04876866, NRC ID: CC-BR-118-1001/1004 Walkdowns requirements not met 2. IR 04872639, NRC ID - Flex Pumps less than 90% fuel level

3. IR 04876831, NRC ID CETI: Acceptance Criteria for 0BwOS FX-12

Performance Assessment:

Performance Deficiency: The failure to implement and maintain BDBEE mitigation strategies and ensure equipment relied on for the mitigation strategies had sufficient capacity and capability to perform their required functions was contrary to 10 CFR 50.155 and a performance deficiency. Specifically, the licensee failed to:

1. Validate and document FLEX strategies were capable of being implemented site-wide

and remained viable on a recurring basis.

2. Maintain availability of the minimum required number of low and medium head FLEX

pumps.

3. Establish flowrate acceptance criteria for the low head FLEX pumps to ensure the

pumps had sufficient capacity and capability to perform their required function.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors 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 failure to implement and maintain BDBEE mitigation strategies and confirm the required equipment had sufficient capacity and capability, did not ensure the site was prepared to effectively maintain or restore core cooling, containment, and spent fuel pool cooling capabilities to prevent undesirable consequences.

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 the finding was of very low safety significance (Green) because it was credited for satisfying EA-12-049 and did not represent an exposure period of greater than 21 days. Therefore, the inspectors answered No to both Exhibit 2, Mitigating Systems Screening Questions, Section E, Flexible Coping Strategies (FLEX) screening questions.

Cross-Cutting Aspect: H.7 - Documentation: The organization creates and maintains complete, accurate and up-to-date documentation. Specifically, the site did not maintain the appropriate documentation and ensure FLEX records were readily accessible. This included retaining FLEX walkdown validations, ensuring surveillance as-left FLEX pump conditions supported the FLEX strategy, and surveillance procedure acceptance criteria reflected the most current, and limiting, design conditions.

Enforcement:

Violation: Title 10 CFR 50.155(b)(1)(i), Mitigation of beyond-design-basis events: Strategies and guidelines, states, in part, each licensee shall develop, implement, and maintain:

Mitigation strategies for beyond-design basis external events - 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 either a loss of normal access to the ultimate heat sink or, for passive reactor designs, a loss of normal access to the normal heat sink.

These strategies and guidelines must be capable of being implemented site-wide and must include the following: Maintaining or restoring core cooling, containment, and spent fuel pool cooling capabilities.

Title 10 CFR 50.155(c), Equipment, states, in part,

(1) The equipment relied on for the mitigation strategies and guidelines required by paragraph (b)(1) of this section must have sufficient capacity and capability to perform the functions required by paragraph (b)(1) of this section.

Procedure CC-BR-118-1004, Braidwood Station Unit 1 & 2 Final Integrated Plan Document, Revision 1, outlines the sites strategies and guidelines for the mitigation strategies for beyond-design basis external events. Procedure CC-BR-118-1001, Site Implementation of Diverse and Flexible Coping Strategies (FLEX) and Spent Fuel Pool Instrumentation Program, Revision 9, describes the program for FLEX strategies at Braidwood. These procedures are used to comply with 10 CFR 50.155(b)(1)(i).

Procedure 0BwOS FX-12, FLEX Pump Full Flow Test Surveillance, Revision 3, performed full flow tests of the FLEX pumps to ensure the pumps had sufficient capacity and capability to perform their required functions. This procedure was used to comply with 10 CFR 50.155(c)(1).

Contrary to the above, as of June 26, 2025, the licensee failed to implement and maintain strategies and guidelines to mitigate beyond-design basis external events as required by 10 CFR 50.155(b)(1)(i) and failed to ensure equipment relied on for the mitigation strategies and guidelines had sufficient capacity and capability to perform the functions as required by 10 CFR 50.155(c)(1).

Specifically, the licensee failed to:

1. Perform and document the sites validation of FLEX strategies as required by

Procedure CC-BR-118-1004, Section 2.17 and 2.18.1 and implemented by Procedure CC-BR-118-1001, Section 4.1.3.

2. Maintain availability of the minimum required low and medium head FLEX pumps as

required by Procedure CC-BR-118-1004, Section 2.3.10.1, CC-BR-118-1001, 9, and 0BwOS FX-1a, Attachment 2.

3. Establish flowrate acceptance criteria, in accordance with Procedure

CC-BR-118-1004, Section 2.3.10.1, for the low head FLEX pump Surveillance Procedure 0BwOS FX-12, Section G, Acceptance Criteria, which ensured equipment had sufficient capacity and capability to perform their required functions.

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

Failure to Translate the Maximum Component Cooling Water Temperature into Transfer to Cold Leg Recirculation Emergency Operating Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-03 Open/Closed

[P.2] -

Evaluation 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to update Emergency Operating Procedures (EOP)1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, to reflect the new maximum component cooling (CC) water temperature following a design-basis loss-of-coolant-accident (LOCA). The new maximum CC temperature was a result of a License Amendment Request (LAR), approved on July 26, 2016, which increased the maximum ultimate heat sink (UHS)temperature.

Description:

Braidwoods component cooling (CC) system was a shared system which provided a heat sink for the removal of process and operating heat from safety-related components during a design-basis accident (DBA) or transient. During normal operation, the CC system also provided this function for various nonessential components as well as the spent fuel pool. The CC system served as a barrier to the release of radioactive byproducts between potentially radioactive systems and the essential service water (SX) system. Each unit had one CC heat exchanger (HX) and one common CC HX was capable of being aligned to either unit such as for a maintenance spare or for additional heat loads. Each HX was sized to be 100 percent capacity to meet the single failure criterion.

The SX system circulated water on the tube-side of the CC HXs with CC water circulated on the shell side. The SX system cooling flow to the CC HXs was regulated by the tube-side outlet motor operated valves, 0/1/2SX007. The ultimate heat sink (UHS) provided a heat sink to the SX system for all safety equipment cooled by SX which were needed for accident mitigation of one unit and simultaneous shutdown/cooldown of the other unit. The limiting DBA for maximum SX heat removal was one unit undergoing a post loss-of-coolant-accident (LOCA) and loss-of-off-site power (LOOP) with the second unit undergoing a safe non-accident shutdown. Under this case, the SX accident unit heat loads, in part, included two CC HXs, emergency diesel generator (EDG) coolers, and reactor containment fan coolers (RCFC). The SX system had two 100 percent capacity pumps, and given a single failure of one SX pump, the remaining pump would be required to cool all associated heat loads that are described in updated final safety analyses report (UFSAR) Table 9.2-1, Essential Service Water Heat Loads.

Braidwood submitted LAR, RS-12-193, Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink, (herein referred to as the UHS LAR) on August 19, 2014 (Agencywide Documents Access and Management System [ADAMS] Accession Number [No.] ML14231A902). This LAR increased the Technical Specification (TS) Surveillance Requirement (SR) 3.7.9.2 UHS allowable average temperature from less than or equal to 100 degrees Fahrenheit (°F) to 102 °F. The purpose of the UHS TS temperature limit was to restrict the initial UHS temperature such that the maximum UHS temperature (i.e., the temperature of the SX cooling water supplied to safety systems) experienced during a DBA would not exceed the design limit of the plant equipment cooled by SX. Due to the increase in UHS temperature, the maximum allowed CC water HX outlet temperature increased from 120°F to 128°F for the accident and non-accident unit during a DBA. This LAR was subsequently approved by the NRC on July 26, 2016 (ADAMS Accession No. ML16133A438).

The inspectors reviewed UFSAR Section 9.2.2.3.5, Post-LOCA Recovery, which stated, in part, During post-LOCA operation, the CC system supply water temperature for the LOCA unit and the non-LOCA unit may increase to 128 °F. UFSAR Table 9.2-3, Component Cooling System Design Parameters, stated, in part, the component cooling water supply temperature (normal operation/cooldown/post-LOCA recovery) °F was 105/120/128 (max).

The inspectors also reviewed EOP 1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 304, Step 11.b.2 to align the CC system for post-LOCA recovery which stated:

Adjust SX007 outlet flow control valve(s) as required to maintain the following conditions:

5. SX pump motor amps - LESS THAN 191 AMPS ON BOTH UNITS

The inspectors questioned why Step 11.b.2 temperature was 120 °F and lower than the UFSAR CC maximum design temperature of 128 °F. As a result, the licensee provided corrective action document (AR) 04867564, Additional actions to support EC 396478 implementation, which was generated prior to the NRC inspection on May 21, 2025. This AR identified the new maximum CC HX outlet temperature of 128°F was not translated into EOPs 1/2 BwEP ES-1.3, Step 11.b.2 at the time the UHS LAR was approved in 2016. The sites recommended actions were to revise the temperature in the EOPs and evaluate if the setpoint calculation needed to be revised. This AR was categorized as a non-corrective action document (NCAP) and therefore, shift review was not required. EOPs 1/2BwEP ES-1.3, Revision 304, was subsequently updated during the inspection on June 11, 2025, to reflect the 128°F temperature in Revision 305. The inspectors also reviewed Engineering Change (EC) 396478, Support Analyses for the License Amendment Request to Raise the Maximum Temperature for the UHS in TS LCO [limiting condition of operation] 3.7.9, Revision 2, which supported Braidwoods UHS LAR. The inspectors noted the site did not recognize the impact to Procedures 1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 300 - 304, in the ECs procedure impact review (Form CC-AA-102-F-09) section.

On June 13, 2025, the inspectors questioned the past impact of the lower CC 120°F temperature on the SX system capability to provide adequate cooling flow to maintain a lower CC temperature post-LOCA from July 26, 2016, to June 11, 2025. This aspect was not recognized by the site in AR 04867564. As a result of the inspectors questions, on June 20, 2025, the site initiated AR 04874976 to evaluate the impact of the lower CC temperature. The immediate operability determination stated the CC system was operable based on the more conservative temperature limitation of 120°F. The inspectors questioned if this was conservative in regard to the SX system and noted more SX flow would be required to maintain a lower CC system temperature. The inspectors additionally questioned what this impact would be on the other SX safety-related heat loads that required cooling post-LOCA and if the SX system capacity was adequate given a single failure of one SX pump (documented in NCV-04 and NCV-05 of this report). Due to the inadequacies of the licensees evaluation of this issue in AR 04867564, this finding was categorized as NRC-identified in accordance with inspection manual chapter (IMC) 0612, Issue Screening.

Corrective Actions: On June 11, 2025, as a result of AR 04867564, the site issued EOPs 1/2BwEP ES-1.3, Revision 305, which reflected the current maximum CC design temperature of 128°F in Step 11.b.2.

On July 23, 2025, under AR 04874976, the site completed, and provided the inspectors, past-operability evaluation and associated EC 645071, Evaluate Past-Operability for CC HX Outlet Temperature Limit of 120°F vs. 128°F in 1/2BwEP ES-1.3, Revision 0. Based on this past-operability evaluation the licensee determined the SX system was operable from approval of the UHS LAR on July 26, 2016, to June 11, 2025, when EOPs Revision 305 was issued. In this evaluation, the site considered the limiting scenario of a LOOP, LOCA, single failure of one SX pump, and the maximum UHS temperature experienced during this time frame. Under this case, one SX pump would be required to provide sufficient cooling flow to maintain the CC outlet temperature to 120°F with the maximum amount of heat removal due to the common HX aligned to the accident unit. The site calculated the total SX flow to both CC HXs to be 17,200 gallons per minute (gpm) with a total combined SX cooling flow for the accident unit, including all safety-related heat loads (i.e., EDGs, RCFCs, etc.), of 33,444 gpm.

This total required flow was below the pump runout flow of 35,000 gpm.

Corrective Action References: AR 04867564, Additional actions to support EC 396478 implementation; AR 04874976, NRC ID CETI: IR 4867564 120F vs 128F - Add Action Needed.

Performance Assessment:

Performance Deficiency: The failure to correctly translate the CC maximum design temperature into EOPs was contrary to 10 CFR 50 Appendix B, Criterion III, Design Control, and a performance deficiency. Specifically, the licensee failed to update EOPs 1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 300 to 304, Step 11.b.2 for the CC maximum design temperature following approval of a LAR to increase the UHS temperature.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality 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 failure to update EOPs to the new maximum CC temperature would have caused the operators to divert SX flow from other safety-related components during an accident to maintain the lower CC temperature. This would have caused the site to operate outside of the analyzed design for the SX system.

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 this finding was of very low safety significance (Green) because past-operability review and associated EC 645071 determined the SX system maintained its operability under the most limiting post-LOCA conditions from July 26, 2016, to June 11, 2025. Therefore, the inspectors answered 'YES' to Question 1 under Inspection Manual Chapter 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions."

Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, during the 2016 procedure impact review under EC 396478, the site failed to evaluate the impact the new maximum UHS temperature had on all EOPs.

Additionally, once the procedure impact was identified under AR 04867564 in 2025, the site failed to evaluate the past impact the lower maximum CC temperature would have had on the associated systems had an accident occurred.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

UFSAR Section 9.2.2.3.5, Post-LOCA Recovery, states, in part, During post-LOCA operation, the CC system supply water temperature for the LOCA unit and the non-LOCA unit may increase to 128 °F. UFSAR Table 9.2-3, Component Cooling System Design Parameters, stated, in part, the component cooling water supply temperature (normal operation/cooldown/post-LOCA recovery) °F was 105/120/128 (max).

EOP 1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 300 to 304, Step 11.b.2 stated:

Adjust SX007 outlet flow control valve(s) as required to maintain the following conditions:

5. SX pump motor amps - LESS THAN 191 AMPS ON BOTH UNITS

LAR, RS-12-193, Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink, was approved on July 26, 2016 (ADAMS Accession No. ML16133A438). This LAR increased the maximum UHS to 102 °F and consequently increased the CC maximum temperature to 128 °F following a design-basis LOCA.

Contrary to the above, from July 26, 2016 to June 11, 2025, the licensee failed to assure that applicable regulatory requirements and the design basis for the safety-related SX and CC systems were correctly translated into specifications, drawings, procedures, and instructions.

Specifically, the post-LOCA CC maximum temperature of 128°F as described in the UFSAR, was not translated into EOPs 1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 300 to 304, after LAR RS-12-193 was approved on July 26, 2016.

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

Failure to Verify the Adequacy of the Essential Service Water System Design Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-04 Open/Closed

[H.6] - Design Margins 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to verify or check the adequacy of the essential service water (SX)system design to ensure all safety-related loads could be adequately cooled during accident conditions. Specifically, the site failed to establish design control measures such as SX system flow balance test procedures, hydraulic calculations, or alternate methods to verify the design capability of the Unit 1 and 2 SX system.

Description:

Braidwood Unit 1 and 2s essential service water (SX) systems were designed to ensure sufficient cooling capacity was available to cool safety-related heat loads during normal, loss-of-coolant-accident (LOCA), loss of offsite power (LOOP), and normal shutdown conditions.

These loads, which were essential to the safe shutdown of the reactor, included, in part, the component cooling (CC) heat exchangers (HX), emergency diesel generator (EDG) coolers, and reactor containment fan coolers (RCFC). Each unit had one CC HX, and a common CC HX was available to be aligned to either unit.

Each units SX system had two full-capacity SX pumps rated at 24,000 gallons per minute (gpm). Each units SX systems were divided into two redundant loops that were capable of being operated separately or cross-tied. In the event of a total loss of all SX on one unit (i.e.,

a beyond-design-basis [BDB] event), Unit 1 and 2 SX systems were capable of being cross tied through the 1/2SX005 motor operated valves. Unit 1 and 2 SX systems had a common suction supply/discharge which was from/to the safety-related, category I essential cooling pond (i.e., the Ultimate Heat Sink [UHS]). Average UHS temperature was required to be less than or equal to 102.8°F (degrees Fahrenheit) until September 30, 2025, and less than or equal to 102°F after September 30, 2025, per Technical Specification (TS) Surveillance Requirement (SR) 3.7.9.2.

On June 13, 2025, the inspectors requested the SX system flow balance surveillance procedure which ensured adequate SX flow was provided to safety-related components during normal, accident, and shutdown conditions. These heat loads and nominal flow rates were described in updated final safety analysis report (UFSAR), Table 9.2-1, Essential Service Water Heat Loads, and Table 9.2-11, Essential Service Water Component Nominal Design Flow Rates. In response, on June 18, 2025, the site provided two surveillance procedures: 1BwOSR 3.6.6.2, Reactor Containment Fan Cooler Surveillance, Revision 31, and 1BwOSR 3.6.6.3-1, SX System Flow Balance Surveillance, Revision 15.

The inspectors reviewed 1BwOSR 3.6.6.2 and noted Section D. Precautions, stated in part, SX flow adjustments are NOT permitted in this procedure. If SX flow balance adjustments to the RCFCs must be performed to meet acceptance criteria, perform 1BwOSR 3.6.6.3-1. The acceptance criteria in procedure 1BwOSR 3.6.6.2 ensured TS SR 3.6.6.2 and 3.6.6.3 were met which verified operability of each RCFC. The inspectors then reviewed procedure 1BwOSR 3.6.6.3-1 and noted the purpose was to flow balance the SX system, however, Section G. Acceptance Criteria, stated, None. On June 18, 2025, the inspectors questioned how the site ensured all safety-related loads received adequate SX cooling flow and noted the most limiting operating conditions for SX heat removal post-LOCA. Specifically, if there were no acceptance criteria in the SX flow balance procedure, how did the as-left SX throttle valve positions for each safety-related load ensure adequate cooling flow would be provided under the limiting post-LOCA operating conditions. The inspectors noted limiting post-LOCA operating conditions such as higher required SX flows through the CC HXs, decreased UHS water level, minimum allowable SX pump performance, SX pump speed adjustment due to minimum allowable EDG frequency variations, and SX strainer differential pressure. The inspectors also questioned if the site had a SX hydraulic system design analysis which ensured adequate cooling to all SX safety loads.

The licensee was unable to provide any additional testing procedures, design reviews, calculations or alternative methods to answer the inspectors questions. As a result, the licensee performed SX flow analyses and on June 26, 2025, documented these analyses in their corrective action program under AR 04876829. The inspectors reviewed the AR and noted the analyses assumed a LOOP with the failure of one EDG such that one train of engineered safety features (ESF) equipment would not operate. The inspectors determined this did not consider the most limiting SX failure of one SX pump in which the remaining SX pump would be required to remove the heat from both ESF divisions (i.e., two trains of residual heat removal [RHR]). Under this more limiting case, the inspectors noted the operators would be expected to direct significantly more SX flow to the CC HXs to maintain CC temperature under 128°F in accordance with emergency operating procedure (EOP)1/2BwEP ES-1.3, Revision 305, Step 11.b.2. Therefore, the inspectors determined the evaluation documented in AR 04876829 did not fully address the concern.

As a result, on July 9, 2025, the site entered AR 04880091 into their corrective action program and on July 18, 2025, completed operability evaluation 25-002 and associated engineering change (EC) 645015. This operability evaluation and EC incorporated the single failure of one SX pump and verified that under these limiting post-LOCA conditions adequate SX flow would be provided to all the supported safety-related components.

Based on their review the inspectors determined the licensee failed to verify the adequacy of the Unit 1 and 2 SX system design to ensure all safety-related heat loads are adequately cooled during accident conditions. The failure to evaluate the single failure of only one SX pump is documented and assessed separately in this inspection report as NCV-05.

Corrective Actions: AR 04876829, Assignments 2-3, planned to evaluate changes to formally record the flow rates to the SX components under surveillance procedures 1/2BwOSR 3.6.6.3-1. Additional assignments included completing an EC to formally document the evaluation of SX flows to safety-related components based on the SX flow balance results.

AR 04880091, Assignment 2, performed operability evaluation 25-002 and directly answered the inspectors concerns since it considered the limiting scenario of a LOCA with a LOOP given a single failure of one SX pump. Under this case, two ESF trains remained in operation with two RHR HXs aligned for post-LOCA cold leg recirculation. With the common CC HX aligned to the accident unit, the maximum amount of SX flow was required for post-LOCA heat removal. Specifically, when one SX pump was supplying both CC HXs, the evaluation verified the impact of higher SX pump flow from a single pump, subsequent lower SX pump discharge pressure and lower SX flows to equipment cooled by SX. The evaluation determined the SX system remained operable since the total required SX flow for the LOCA unit, about 30,000 gpm, was below the SX pump runout flow of 35,000 gpm. Additionally, the evaluation determined all equipment cooled by SX (e.g., CC HXs, RCFCs, EDG coolers)remained operable with the calculated reduced SX flows.

As a result of the operability evaluation, AR 04880091 documented additional corrective action assignments which included: developing an SX hydraulic model to evaluate the flow path for the BDB SX unit cross-tie when only one SX pump would be running on the LOCA unit and two ESF trains were in operation. The site planned to review options and develop the required actions to restore required flow rates to the equipment supplied by SX. Based on the results of these actions, the site planned to incorporate the BDB SX unit cross-tie into the Braidwood design basis or implement an option to restore required flow rates. Additionally, the site planned to create a design analysis to document post-LOCA SX flow rates or revise the flow balance procedure (1/2BwOSR 3.6.6.3-1).

Corrective Action References: AR 04876829, "NRC ID CETI: SX Flow to Safety Related Components post-LOCA; AR 04880091, NRC ID CETI: SX Pump Bounding Single Failure

Analysis.

Performance Assessment:

Performance Deficiency: The failure to verify the adequacy of the Unit 1 and 2 SX system design to ensure all safety-related heat loads are adequately cooled during accident conditions was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and performance deficiency. Specifically, the licensees SX flow balance procedure, 1BwOSR 3.6.6.3-1, Revision 15, did not establish acceptance criteria for all required SX heat loads.

Alternatively, the licensee did not have additional testing procedures, design reviews, calculational, or other alternative methods to ensure all design bases SX flows would be achieved under all operating conditions.

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. This issue was more than minor due to the potential impact, during accident conditions, to the SX system and safety-related components cooled by SX.

Specifically, the EOPs do not provide actions to adjust all SX supplied heat loads, such as the EDG coolers. The flow balance procedure does not ensure the as-left position of these valves provide adequate SX cooling flow under limiting post-LOCA conditions. Alternatively, the site did not have a design analysis to ensure the SX system was capable of providing adequate SX cooling flow to all heat loads which were essential to the safe shutdown of the reactor.

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 this finding was of very low safety significance (Green) because operability evaluation 25-002 determined the SX system, and all associated equipment cooled by SX, maintained its operability under the most limiting post-LOCA conditions (i.e.,

single failure of an SX pump, post-LOCA). Therefore, the inspectors answered YES to Question 1 under Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions.

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, in the last three years, the licensee submitted multiple License Amendment Requests for TS SR 3.7.9.2 which increased the UHS temperature. An increase in UHS temperature could affect the required SX flow needed for each SX safety-related heat load. In the process of performing these changes, if the licensee used a systematic and rigorous approach to guard margins, the licensee could have recognized this decrease in SX design margins and identified they did not have a design analysis and/or test to verify sufficient SX flow remained available to all SX heat loads.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that the licensee provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

UFSAR Section 9.2.1.2.1, Essential Service Water System: Design Bases, stated, in part, The essential service water system is designed to ensure that sufficient cooling capacity is available to provide adequate cooling during normal and accident conditions. The components served by the essential service water for normal, LOCA, loss of offsite power (LOOP), or shutdown conditions are shown in Table 9.2-1. UFSAR Table 9.2-1, Essential Service Water Heat Loads, listed the SX heat loads during normal, LOCA, LOOP, and shutdown conditions. These heat loads included, in part, the EDG coolers, RCFCs, and CC HXs. UFSAR Table 9.2-11, Essential Service Water Component Nominal Design Flow Rates, listed the nominal flow (gpm) required for each SX heat load listed in Table 9.2-1.

Surveillance Procedure, 1/2BwOSR 3.6.6.2, Reactor Containment Fan Cooler Surveillance, Revision 31, verified the RCFC operability in Modes 1, 2, 3, and 4 in accordance with T.S.

S.R. 3.6.6.2 and 3.6.6.3. Surveillance Procedure, 1/2BwOSR 3.6.6.3-1, SX System Flow Balance Surveillance, Revision 15, performed the flow balance of the SX system. Section G, Acceptance Criteria, stated none.

Contrary to the above, as of July 28, 2025, the licensee failed to verify the adequacy of the Unit 1 and 2 safety-related SX system design. Specifically, the SX system capability to perform its design function under accident conditions including the limiting post-LOCA operating conditions and postulated single failure of one SX pump was not verified by hydraulic design reviews, calculational methods or by an adequate testing program.

Surveillance Procedure, 1BwOSR 3.6.6.3-1, SX System Flow Balance Surveillance, Revision 15, had no acceptance criteria provided in Section G, Acceptance Criteria, to verify all SX heat loads could be adequately cooled under normal, accident, and shutdown conditions. Alternatively, the licensee did not have additional design reviews, calculational or alternative methods to verify the adequacy of the SX system design.

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

Failure to Evaluate a Single Failure of an Essential Service Water Pump Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-05 Open/Closed

[H.14] -

Conservative Bias 71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to translate a limiting single failure of the essential service water (SX)system under post loss-of-coolant-accident (LOCA) conditions into specifications, drawings, procedures, and instructions design documents. Specifically, the licensee failed to assure a single SX pump on the accident unit was capable of providing sufficient cooling to all required SX heat loads accident conditions.

Description:

Braidwoods essential service water (SX) system was designed to ensure sufficient cooling capacity was available to cool safety-related heat loads during normal, accident, and shutdown conditions. Each unit had two redundant SX loops that were capable of being operated separately or cross tied. Each unit also had two full-capacity SX pumps rated at 24,000 gallons per minute (gpm). The components cooled by the SX system under various operating conditions were described in updated final safety analysis report (UFSAR) Table 9.2-1, Essential Service Water Heat Loads. The heat loads during a loss-of-coolant-accident (LOCA), in part, included the component cooling (CC) heat exchangers (HX), emergency diesel generator (EDG) coolers, and reactor containment fan coolers (RCFC). UFSAR Table 9.2-2, Single-Failure Analysis of the Essential Service Water System, identified a failure of an SX pump (1SX01PA/B, 2SX01PA/B) to operate. The tables alternative column noted there were two full-capacity pumps available in each unit.

The inspectors reviewed the emergency operating procedures (EOP) strategy to mitigate a LOCA. Specifically, the inspectors reviewed EOP 1/2 BwEP-0, Reactor Trip or Safety Injection, Revision 307, and EOP 1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 305. EOP 1/2 BwEP-0, Step 12.a verified both SX pumps were running. The response not obtained column had an action to manually start the pump(s). Attachment B, Balance of Plant Verification, Step 4, adjusted the required SX flow by throttling the CC HX outlet motor operated valves (MOV), 0/1/2SX007, based on the number of SX pumps running on each unit and the number of CC HXs in service. Each unit had one CC HX and a common HX was available to be aligned to either unit.

EOP 1/2BwEP ES-1.3, Step 11, aligned the CC system for post-LOCA recovery. This included checking SX flow to the common unit CC HX (Step 11.a) and adjusting SX flow by throttling the 0/1/2SX007 outlet MOVs to maintain the following conditions (Step 11.b.2):

a.

CC HX outlet temperature - LESS THAN 128 °F [degrees Fahrenheit]

b.

SX pump motor amps - LESS THAN 191 AMPS ON BOTH UNITS On June 13, 2025, the inspectors questioned the post-LOCA capability of the SX system.

Specifically, the inspectors postulated a single failure of one SX pump (UFSAR Table 9.2-2)on the accident unit and questioned if the remaining SX pump had the capacity to provide the required cooling flow to all SX heat loads (UFSAR Table 9.2-1). Under these conditions, the inspectors noted the operators would be directed to provide significantly more flow to the CC HXs in EOP 1/2BwEP ES-1.3, Step 11.b.2, and questioned if one SX pump would be capable of providing adequate CC cooling flow while ensuring SX flow would not be diverted from other heat loads (i.e., EDG coolers, RCFCs). On June 26, 2025, the site addressed one aspect of the inspectors concerns which regarded flow balancing the SX system under AR 04876829 (documented in NCV-04 of this report). However, the single failure of one SX pump was not included in this evaluation. On June 27, 2025, the site, including operations staff, discussed the single failure of one SX pump with the inspectors. In order to address the immediate operability of the SX system, the site provided a beyond-design-bases (BDB)approach which allowed Unit 1 and 2 SX systems to be cross tied. However, the site did not provide an approach to mitigate this scenario that was within their design bases. A written response describing this BDB approach was provided to the inspection team on July 2, 2025.

The licensee was unable to provide existing design documents that supported one SX pump's capability of providing the required SX cooling flow to all SX heat loads and on July 9, 2025, entered AR 04880091 into their corrective action program. In the AR, the site noted their current design basis analysis for containment integrity post-LOCA assumed a LOOP with the failure of one EDG such that one train of engineered safety features (ESF)equipment did not operate. While these assumptions were conservative to determine the maximum containment response, they were not conservative with regard to the SX system.

Specifically, the inspectors were concerned that if one SX pump failed, would the remaining SX pump be capable of adequately cooling all safety-related heat loads, including both ESF trains in EOP 1/2BwEP ES-1.3, Step 11.b.2.

Corrective Actions: On July 18, 2025, the site completed, and provided to the inspectors, operability evaluation 25-002 and associated EC 645015 under AR 04880091, Assignment 2.

Based on this operability evaluation the licensee determined the SX system remained operable. In this evaluation, the site considered the limiting scenario of a LOCA with a LOOP and single failure of one SX pump. Under this case, one SX pump would be required to provide sufficient cooling flow for the heat removal of two ESF trains with two residual heat removal (RHR) HXs aligned during post-LOCA cold leg recirculation. The site calculated the SX flow to each heat load would be reduced by 13 percent and each component was still capable of performing their intended safety functions. The combined SX flow under this condition was calculated to be approximately 30,000 gpm which was below the SX pump runout flow of 35,000 gpm.

Additional corrective action assignments included: developing a SX hydraulic model to evaluate the flow path for the BDB SX unit cross tie when only one SX pump was available on the LOCA unit and two ESF trains were in operation. The site planned to review options and develop the required actions to restore required flow rates to the equipment supplied by SX. Based on the results of these actions, the site planned to either incorporate the BDB SX unit cross tie into the Braidwood design basis or implement an option to restore required flow rates. Additionally, the site planned to create a design analysis to document post-LOCA SX flow rates or revise the flow balance procedure (1/2BwOSR 3.6.6.3-1).

Corrective Action References: AR 04880091, NRC ID CETI: SX Pump Bounding Single Failure

Analysis.

Performance Assessment:

Performance Deficiency: The failure to translate the limiting single failure of the SX system into specifications, drawings, procedures, and instructions, was contrary to Criterion III, Design Control, and was a performance deficiency. Specifically, the licensee did not translate and evaluate the capability of one SX pump on Unit 1 or 2 to provide sufficient cooling to all required SX heat loads during accident conditions given a single failure of one SX pump into specifications, drawings, procedures and instructions.

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 failure to analyze a single failure of one SX pump did not ensure adequate SX cooling flow would be available to all required safety-related loads including the CC HXs, EDG coolers, and RCFCs. Since EOP 1/2BwEP ES-1.3, Step 11.b.2, directed operators to control SX flow based on the CC outlet temperature, under a single failure of one SX pump, the site did not analyze the SX system capability to perform this function. The site only had BDB coping mechanisms available (i.e., Unit 1 and 2 SX system cross tie).

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 this finding was of very low safety significance (Green) because operability evaluation 25-002 determined the SX system maintained its operability under the most limiting post-LOCA conditions given a single failure of one SX pump. Therefore, the inspectors answered 'YES' to Question 1 under Inspection Manual Chapter 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions."

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee's original evaluation under AR 04876829 and current design analysis for containment integrity post-LOCA assumed a single failure of an ESF train was the most limiting and conservative single failure. While this was accurate for containment response, it was not bounding with regards to the capacity of the SX system (documented in AR 04880091).

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

UFSAR Table 9.2-1, Single Failure Analysis of the Essential Service Water System, identified a failure of one SX pump (1SX01PA/B, 2SX01PA/B). The table's alternative column stated, two full-capacity pumps available in each unit. UFSAR Table 9.2-1, Essential Service Water Heat Loads, listed the SX heat loads during normal, LOCA, LOOP, and shutdown conditions. UFSAR Table 9.2-11, Essential Service Water Component Nominal Design Flow Rates, listed the nominal flow (gpm) required to each SX heat load.

EOP 1/2BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 305, Step 11.b.2 stated:

Adjust SX007 outlet flow control valve(s) as required to maintain the following conditions:

7. SX pump motor amps - LESS THAN 191 AMPS ON BOTH UNITS

Contrary to the above, as of July 28, 2025, the licensee failed to assure that applicable regulatory requirements and the design basis for the safety-related SX system were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to ensure the Unit 1 and 2 SX system design, including a single failure of one SX pump on either unit as described in the UFSAR, was translated into specifications, drawings, procedures and instructions. During a post-LOCA, the site did not assure EOP 1/2BwEP ES-1.3, Step 11.b.2, to maintain a CC temperature less than 128°F could be accomplished when only one SX pump was available while also ensuring all other required SX heat loads described in UFSAR Table 9.2-1 could be adequately cooled.

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

Failure to Identify Local Operation of MOV 1(2)CC9415, After a Loss of Electrical Power, was Within Design Bases Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000456,05000457/2025010-06 Open/Closed None (NPP)71111.21M The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.55a(f)(4)(ii), Applicable IST Code: Successive code of record intervals, when the licensee failed to identify local operation to close MOV 1(2)CC9415, using the handwheel, was required to comply with the site's design bases. Specifically, following certain design bases accidents or transients combined with a single failure of the electrical power supply to 1(2)CC9415, operators would need to locally close 1(2)CC9415 to isolate the non-essential (service loop) loads of the CC system and ensure sufficient CC cooling flow remains available to the essential (safety-related) loads. Additionally, the licensees Inservice Testing (IST) program failed to include the full stroke tests required for credited manual valves.

Description:

The safety-related Component Cooling Water (CC) System is a shared system which provides a heat sink for the removal of process and operating heat from safety-related components during a Design Basis Accident (DBA) or transient. During normal operation, the CC System also provides this function for various nonessential components, as well as the spent fuel pool (SFP). The nonessential components and SFP are supplied by the service loop portion of the CC System. Valves 1(2)CC9415, one per Unit, are the CC U-1(2) Service Loop Isolation valves. These are Motor Operated Valves (MOVs), and as the name implies, are the credited valves used to isolate the nonessential loads of the CC system from the remaining safety-related loads. This load shedding ensures sufficient CC cooling flow is available to the required loads in the accident analyses (e.g., the Residual Heat Removal Heat Exchangers). This is discussed in the Updated Final Safety Analysis Report (UFSAR)

Table 9.2-4, Typical/Nominal System Flow Conditions for Main Plant Operating Phases (One Unit), which include the following note:

Note

(8) CC flow to the Service Loop may be isolated by closing the CC9415 valve to ensure the required CC flow through each residual heat removal heat exchanger and to prevent CC pump run out during the recirculation phase of a LOCA.

As part of the selected inspection samples, the team reviewed 10 CFR 50.59 screening BRW-S-2023-047, Revision 0, Breaker Position Change for MOVs 1(2)CC9415, to Eliminate Spurious Closure. The team initially questioned if a time critical operator action (TCA) should have been created for the new operator field action to reenergize the valve prior to closing it during an accident. Further investigation determined that accident analyses assume this valve is closed prior to initiating containment sump cooling after a DBA LOCA.

On June 12, 2025, the inspectors determined if power to 1(2)CC9415 was lost due to a single failure of its electrical power supply, 1(2)CC9415 would have to be closed locally using its handwheel. Of particular concern, a single failures of either vital power bus would result in power loss to one of the 1(2)CC9415 valves and at least one of the associated CC pumps.

The licensee had not previously recognized this single failure combined with local valve closure was within the site's design bases, as this valve manipulation was the sites de facto credited approach to mitigate some DBAs or transients. The inspectors were concerned because the time delay associated with locally closing the valve, using the handwheel, after a DBA was not considered in accident analyses and there was no TCA to validate the time required to complete this action. Later this same day, the inspectors notified the licensee of their concern. The licensee stated they were arriving to the same conclusion based on the conversations and questions from the inspectors the previous days.

The inspectors also noted, and discussed with the licensee, the 1(2)CC9415 are included in the sites Inservice testing (IST) program. However, the local handwheel function of the valve was not being periodically tested as required by 10 CFR 50.55a(f), "Preservice and Inservice testing."

The 2012 American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, is the sites code of record for IST. Paragraph ISTC-3540, Manual Valves, states:

Manual valves shall be full-stroke exercised at least once every 2 yr, except where adverse conditions may require the valve to be tested more frequently to ensure operational readiness. Any increased testing frequency shall be specified by the Owner. The valve shall exhibit the required change of obturator position.

For the purpose of the ISTC-3540 and this writeup, when referring to manual it is meant as locally using the handwheel. This clarification is provided, as sometimes valves operated from the control room are also informally referred to as "manually" by the licensee.

Additionally, the inspectors identified an issue with the licensees conclusion documented in 50.59 screening BRW-S-2023-047, Breaker Position Change for MOVs 1(2)CC9415 to Eliminate Spurious Closure. This issue is documented separately in this report as an Unresolved Item (URI).

Corrective Actions: The licensee initiated multiple corrective action documents to begin evaluating the identified concerns, including:

(1) the need to close 1(2)CC9415 via the handwheel following a single failure of its power supply;

(2) the lack of a TCA associated closure of 1(2)CC9415; and

(3) the lack of manual valve testing as required by the IST.

Additionally, the licensee performed a preliminary evaluation to assess the potential impact associated with the delay due to locally closing 1(2)CC9415 via the handwheel. The licensee estimated the time to close 1(2)CC9415 would increase from about two minutes to over 20-30 minutes. In their preliminary evaluation, the licensee performed a computation to estimate the lower CC flows to the Residual Heat Removal Heat Exchangers during the Emergency Core Cooling System recirculation alignment. This computation assumed a duration of 45 minutes for the lower CC flow condition. The results showed an increase in the Containment Pressure and Containment Temperature but remained below the peak calculated pressure for the LOCA event and the Technical Specifications limit. The increased temperature also remained below the design temperature of the containment liner and internal containment structure. The licensee's evaluation also showed an increase to the containment sump recirculation temperature, which impacted the Net Positive Suction Head (NPSH) current analyses. The licensee evaluated these impacts and concluded the supported systems remained operable.

Corrective Action References: AR 04873558, NRC ID CETI: Legacy Design Issue - Power Source to 1/2CC9415 AR 04875623, NRC ID CETI: MOV 1/2CC9415 Manual handwheel IST Requirement AR 04873820, NRC ID CETI: TCA to Power and Close 1/2CC9415 in LOOP-LOCA

Performance Assessment:

Performance Deficiency: The licensee failed to identify local operation to close MOV 1(2)CC9415, using the handwheel, was required to comply with the site's design bases. This was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, 10 CFR Part 50.55a(f)(4)(ii), Applicable IST Code: Successive code of record intervals, and a performance deficiency. Specifically, following certain design bases accidents or transients, combined with a single failure of the electrical power supply to 1(2)CC9415, operators would need to locally close 1(2)CC9415 to isolate the non-essential (service loop) loads of the CC system and ensure sufficient CC cooling flow remains available to the essential (safety-related) loads.

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, local operation via the handwheel to close MOV 1(2)CC9415 would cause a delay in isolating the non-essential service loop. Preliminary evaluations estimated the delay could significantly increase the time to isolate from the previously assumed two minutes to over 20-30 minutes. This delay would result in reduced CC cooling flow to the residual heat removal (RH) heat exchanger, which post-accident would cause an increase to the calculated containment pressure, containment temperature and containment sump temperature.

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 screened the finding to Green (very low safety significance) because the finding did not result in the loss of operability or PRA functionality. This was based on answering Yes to question A.1 of the Exhibit 2 - Mitigating Systems Screening Questions. Specifically, the licensee's preliminary evaluations determined the impact on the accident analyses did not result in any design limits being exceeded.

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

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures and instructions.

Title 10 CFR Part 50.55a(f)(4)(ii), Applicable IST Code: Successive code of record intervals, requires, in part, Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 18 months before the start of the 120-month interval.

The licensees ASME Code of Record for Operation and Maintenance of Nuclear Power Plants is the 2012 Edition.

The 2012 ASME OM Code, Paragraph ISTC-3540, Manual Valves, states, Manual valves shall be full-stroke exercised at least once every 2 yr, except where adverse conditions may require the valve to be tested more frequently to ensure operational readiness. Any increased testing frequency shall be specified by the Owner. The valve shall exhibit the required change of obturator position.

Braidwood Nuclear Power Station Units 1 & 2 - Inservice Testing Program Fourth Ten Year Interval, Revision 8, is the site's IST program document and establishes the requirements for the performance and administration of assessing the operational readiness of those pumps and valves whose specific functions are required to: (1)Shutdown the reactor to the safe shutdown condition;

(2) Maintaining the safe shutdown condition; or

(3) To mitigate the consequences of an accident.

UFSAR Table 9.2-4, Typical/Nominal System Flow Conditions for Main Plant Operating Phases (One Unit), include note (8), which state, CC flow to the Service Loop may be isolated by closing the CC9415 valve to ensure the required CC flow through each residual heat removal heat exchanger and to prevent CC pump run out during the recirculation phase of a LOCA.

Contrary to the above, as of June 26, 2025, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures and instructions when the licensee did not identify local operation to close MOV 1(2)CC9415, using the handwheel, was required to comply with the site's design bases. Specifically, following certain design bases accidents or transients combined with a single failure of the electrical power supply to 1(2)CC9415, operators would need to locally close 1(2)CC9415 to isolate the non-essential (service loop) loads of the CC system and ensure sufficient CC cooling flow remains available to the essential (safety-related) loads.

In addition, since the function of locally operating 1(2)CC9415 using the handwheel is a function required for safety, Inservice testing to verify operational readiness of 1(2)CC9415 are required. Specifically, 1(2)CC9415 are identified as active valves in the site's IST Program document. However, the full stroke tests required by ISTC-3540 for manual valves were not included in the program's requirements for 1(2)CC9415.

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

Unresolved Item (Open)10 CFR 50.59 Evaluation Needed to Determine if Breaker Position Change for 1(2) CC9415 required NRC Approval URI 05000456,05000457/2025010-07 71111.21M

Description:

On June 11, 2025, the inspectors identified, and notified the licensee, of an issue of concern which resulted in an Unresolved Item (URI) associated with the licensees 50.59 Screening BRW-S-2023-047, Breaker Position Change for MOVs 1(2) CC9415 to Eliminate Spurious Closure, Revision 0. On June 25, 2024, the licensee approved the screening in support of a change to the normal position of the circuit breakers for MOVs 1(2) CC9415 from the normally ON position to OFF under engineering change (EC) 624189. Under EC 624189, the licensee stated a step would be added to licensee procedure 1(2) BwEP-1, Loss of Reactor Coolant or Secondary Coolant, to dispatch an operator to perform actions to close the breaker locally and, if necessary, locally close valve 1(2) CC9415 when directed by the main control room. The licensee highlighted that the added step would ensure that the breaker position change from ON to OFF did not delay actions required to support the component cooling water system required function post-accident and that local operation of valve 1(2)

CC9415 was already part of the sites design basis as documented in procedure 1(2) BwEP ES-1.3. Based on the screening, the licensee determined that the change of breaker position was not adverse, and no 50.59 Evaluation was required.

Valves 1(2)CC9415, one per Unit, are the CC U-1(2) Service Loop Isolation valves. These are Motor Operated Valves (MOVs) used to isolate the nonessential loads of the Component Cooling Water (CC) system from the remaining safety-related loads. This ensures sufficient CC cooling flow is available to the supported safety-related equipment. Use of 1(2)CC9415 is discussed in the Updated Final Safety Analysis Report (UFSAR) Table 9.2-4, Typical/Nominal System Flow Conditions for Main Plant Operating Phases (One Unit),in the following note:

Note

(8) CC flow to the Service Loop may be isolated by closing the CC9415 valve to ensure the required CC flow through each residual heat removal heat exchanger and to prevent CC pump run out during the recirculation phase of a LOCA.

The inspectors reviewed screening BRW-S-2023-047 and concluded the licensee incorrectly determined the proposed change to remove power to valves 1(2) CC9415 via the associated circuit breaker was not adverse. Specifically, by changing the normal position of the circuit breakers for MOVs 1(2) CC9415 from ON to OFF, the proposed change in EC 624189 fundamentally altered the means of controlling the design function of 1(2) CC9415 in that there was an additional step required to be locally performed in the field to allow normal valve operation. Prior to the change, the valve could be operated from the main control room, without the need for field activities. Therefore, the inspectors determined that the failure to perform a 50.59 Evaluation for EC 624189 prior to determining if the change could be made without NRC approval was a performance deficiency and a violation of 10 CFR 50.59.

As an additional note, this inspection report documents a related violation identified during this inspection effort. Specifically, the inspectors identified a single failure which would require the operators to close 1(2)CC9415 using the handwheel. This new identified deficiency would extend the time required to operate valves 1(2) CC9415 longer than the timeframe documented in EC 624189, The information reviewed by the inspectors related to this violation helped informed any potential safety concern associated with the URI.

Planned Closure Actions: The inspectors need the licensee to complete the 50.59 Evaluation.

Once the licensee informs the inspectors the evaluation is complete, the inspectors can review and assess the licensees conclusions and determine the severity and final dispositioning of the violation.

Licensee Actions: The licensee acknowledged the added steps to close the breakers can potentially delay the site's actions and complicate the overall response sequence to the event.

As a result, the licensee concluded the change was adverse and documented the inspectors concerns in the corrective action program. The licensee initiated actions to revise EC 624189, revise BRW-S-2023-047 and write a formal 50.59 evaluation.

As discussed above, the licensee also evaluated the impacts caused by the related violation for the required use of the 1(2)CC9415 handwheel. For this related concern the licensee determine there were impacts due to this additional delay, but the affected equipment remained operable (see related violation for further detail).

Corrective Action References: AR 04873831, NRC ID CETI: Incorrect 50.59 Screening BRW-S-2023-047 AR 04873558, NRC ID CETI: Legacy Design Issue - Power Source to 1/2CC9415

EXIT MEETINGS AND DEBRIEFS

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

• On June 26, 2025, the inspectors presented the Initial Technical Debrief inspection results to Brian Bergmann and other members of the licensee staff.

• On July 28, 2025, the inspectors presented the Final Technical Debrief inspection results to Adam Schuerman and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

2-M-065

Braidwood Units 1 & 2 SX Differential Pressure Calculation

19-AN-7

Protective Relay Settings for 4.16kV ESF Switchgear

19-AQ-68

Division Specific Degraded Voltage Analysis

19-T-6

Diesel Generator Loading during LOOP/LOCA - Braidwood

Units 1 & 2

19AQ-68

Division Specific Voltage Analysis

ATD-0109

Thermal Performance of UHS During Postulated Loss-of-

Coolant-Accident

BRA-1SX007

MIDACALC Results

BRW-02-0032-I

Essential Service Water Discharge Header Temperature

Instrument Indication Uncertainty

0B

BRW-13-0160-M

Flex Pump Sizing and Hydraulic Analysis

BRW-14-0030-M

Godwin Pump Suction Line Hydraulic Analysis to Support

FLEX

BRW-19-0037-E

Thermal Overload Sizing for GL 89-10 Throttle MOVs

BRW-97-1072-M

Component Cooling Heat Exchanger Tube Plugging

Evaluation

BRW-97-1072-M

Component Cooling Heat Exchanger Tube Plugging

Evaluation Key Calc

BRW-97-1072-M

Component Cooling Heat Exchanger Tube Plugging

Evaluation

006 & 007

BRW-97-473-E

25 VDC System Short Circuit Calculation

2B

BRW-97-474-E

25 VDC System Short Circuit Calculation

BRW-97-475-E

25 VDC Fuse Sizing and Coordination

6B

C-1101-730-

E420-014

TMI Thermal Overload Evaluation for Jogging GL89-10

MOVs

CE-BB-003

MOV Seismic and Weak Link Analysis for Butterfly Valves

Design Analysis

050991

Piping Stress Analysis for Subsystem 2SX36

2E

71111.21M

Calculations

Design Analysis

051853

Piping Stress Report for the Essential Service Water System

(Subsystem 1SX02)

005M - P

00687232

Min Wall on SX Piping

10/20/2007

00706376

0SX115F SX Suction Valve Pit Flooded and Overflowing

2/03/2007

00712974

SOER 07-02 Response

09/30/2009

00753038

Enhancement Needed to Support SOER 07-02

Implementation

03/21/2008

244811

0SX115E Gland/Bonnet Bolting and Studs Found Corroded

07/28/2011

2729845

Results of EOC Coating Inspection for 0SX115E

10/19/2016

288124

10-OSP-R 0SX115C Injection Flange Leak

10/15/2019

288743

2-OSP-R 0SX115E Coating Inspection

10/17/2019

289130

0SX115C inspection

10/18/2019

04421889

1E/1F CW Traveling Screens Level Transmitter Bypassed

05/06/2021

04512381

Improvements for FLEX/B.5.b Dry Hydrant Suction Pipe

07/22/2022

04512381

Improvements for FLEX/B.5.b Dry Hydrant Suction pipe

07/22/2022

04544721

1C CW abnormal indications

2/24/2022

04544748

Design limits of 1E/1F Traveling Screens Challanged

2/25/2022

04720989

1L-SW007 Indication Issue

2/01/2023

04731153

Unit 1 Traveling Screens Malfunctioning

01/15/2024

04867564

Additional Actions to Support EC 396478 Implemenation

05/21/2025

2589930

Severe Corrosion and Degraded Coating Observed on Line

0SXH2AA-6

11/19/2015

Corrective Action

Documents

23422

2A Pump, Sulzer Evaluated Discrepancies

11/03/2016

04854904

NRC ID CETI: Minor Oil Weepage from 1SX01PA Bearing

Oil Line

04/09/2025

04872271

NRC ID CETI: UFSAR Ch. 9 Needs Updated for SX Supply

Temp

06/11/2025

04872601

NRC ID CETI: Equipment Inappropriately Staged by SX

Pump Room

06/12/2025

04872639

NRC ID CETI: Flex Pumps Less than 90% Fuel Level

06/12/2025

04872659

NRC ID CETI: CC-BR-118-1001/1004 Discrepancies

06/12/2025

04873520

NRC ID CETI: Revise Calc BRW-16-0001-E

06/16/2025

04873558

NRC ID CETI: Legacy Design Issue - Power Source to

1/2CC9415

06/16/2025

71111.21M

Corrective Action

Documents

Resulting from

Inspection

04873784

NRC ID CETI: Update BRW-S-2023-013 50.59 Screening

References

06/17/2025

04873820

NRC ID CETI: TCA to Power and Close 1/2CC9415 in

LOOP-LOCA

06/17/2025

04873831

NRC ID CETI: Incorrect 50.59 Screening BRW-S-2023-047

06/17/2025

04874747

NRC ID CETI: Charger 112 Potential Low Voltage

06/29/2025

04874888

NRC ID CETI: Errors in WO 01807512-01 (2018)

06/20/2025

04874971

NRC ID CETI: SX 007 MOV Throttling Procedure Gaps

06/20/2025

04874976

NRC ID CETI: IR 4867564 120F vs 128F - Add Action

Needed

06/20/2025

04874978

NRC ID CETI: Document Sensitivity Runs in Calc Revision

06/20/2025

04874982

NRC ID CETI: Unit 2 SX Pump Motor BHP Evaluations

Missing

06/20/2025

04875608

NRC ID CETI: 1A CC Heat Exchanger Inspection Report

Errors

06/23/2025

04875623

NRC ID CETI: MOV 1/2CC9415 Manual Handwheel IST

Requirements

06/23/2025

04875891

NRC ID CETI: Error in Input Reference for BRW-13-0160-M

Calc

06/24/2025

04876049

NRC ID CETI: Errors in WO 05080381 (2024)

06/25/2025

04876829

NRC ID CETI: SX Flow to Safety-Related Components

Post-LOCA

06/26/2025

04876831

NRC ID CETI: Acceptance Criteria for 0BwOS FX-12

06/26/2025

04876860

NRC ID CETI: Discrepancy with PRA Model

06/26/2025

04876866

NRC ID CETI: CC-BR-118-1001/1004 Walkdowns

requirements not met

06/26/2025

Corrective Action

Documents

Resulting from

Inspection

04880091

NRC ID CETI: SX Pump Bounding Single Failure Analysis

07/09/2025

1663D96

Motor Outline 1SX01PA

20E-1-4001A

Station One Line Diagram

v

20E-1-4006A

Key Diagram 4160V ESF SWGR Bus 141 (1AP05E)

20E-1-4010A

Key Diagram 125V DC ESF Distribution Center Bus 111

(1DC05E)

M

20E-1-4030SX27

Schematic Diagram Component Cooling Heat Exchanger

Outlet

L

20E-4010B

Key Diagram 125V DC ESF Distribution Center Bus 111

(1DC05E) Part - 2

J

71111.21M

Drawings

M-42

Diagram of Essential Service Water Units 1 & 2

1A

M-42

Diagram of Essential Service Water Unit 1 & 2

2B

M-42

Diagram of Essential Service Water Units 1 & 2

1B

04785457

Vs-VC Cross-tie Performance Issues

07/08/2024

252064

SX Piping Leakage at the Outlet of U-0 CC HX

2/13/2023

394159

Install Dry Hydrant at the Lake Screen House to Support

FLEX Mod 7

05/11/2015

4494801

SX piping leakage at the outlet of U-0 CC HX

04/22/2022

236515

HX Inspection Report

01/27/2025

2930

DC System Short Circuit Current Calculation Update

26545

Proceduralized TCC - Evaluation of Using VS to Cool the

MCR While One Train Of VC is Out of Service for

Maintenance

2/15/2018

634937

MIN WALL THICKNESS CALC FOR LINES 2SX05CA-6"

AND 2SX04DA-6" FOR UT EXAMINATION

636627

N-513-4 CODE CASE EVALUATION FOR PIN HOLE LEAK

AT WELD ON LINE 0SX03A-30"

639395

Braidwood 2: LR Commitment #10-SG Divider Plate

PWSCC

03/15/2024

641059

Support Analyses for the License Amendment Request to

Raise the Maximum Temperature for the UHS TS 3.7.9

Utilizing a Diurnal Temperature Profile

2840

Disposition of Component Cooling Water Temperature

Increase During Normal Operation

644844

Evaluation of Unit 2 SX Pump Motors Brakehorespower

06/20/2025

DRP 20-048

Align Byron and Braidwood Reference to ASTM Standard

D975-06b in UFSAR Appendix F

01/25/2024

Engineering

Changes

ER-AA-340-1002

SERVICE WATER HEAT EX CHANGER INSPECTION

GUIDE

Engineering

Evaluations

631169

Clarify Testing Criteria for Design Basis Accident Flow Rates

for the Essential Service Water Pumps

50.59 Screening:

BRW-S-2023-013

Revise TRM Appendix M, Technical Specification Bases B3.8.3, and UFSAR Appendix F to Align Byron and

Braidwood Station Diesel Fuel Oil Testing Programs

01/25/2024

50.59 Screening:

BRW-S-2025-018

Retention of PBI 24148 for Turbine Building Louver Tarps to

Remain Beyond Ninety-Days

03/18/2025

Miscellaneous

BB-PBD-AMP-

Program Basis Document BB-PBD-AMP-XI.M20 Open-Cycle 07/20/2015

XI.M20

Cooling Water System

CC-AA-201-F-01

Plant Barrier Impact 24148

2/16/2024

CC-AA-201-F-02

Plant Barrier Impact 24148

2/17/2024

CL-QR-15346

Class 1E Battery Chargers Qualification Report

09/27/1979

CY-BR-120-4130

Braidwood Macrobiological Strategic Plan

DHC 770322-1

Caluclated Data Speed Torque Power Factor

03/22/1977

DHC 770322-4

Safe Operating Time

05/09/1978

DRP 20-048

F:

ER-AA-700-1004

LRR Change / Deletion Checklist

01/25/2024

EC 394159:

HL130M Dri-

Prime Pump

Vendor Manual for Low Head FLEX Pump

Limitorque Corp

Memo

Motor Jogging Capabilities

2/25/1992

Technical

Specification

Change: 23-001

Technical Specification Change Request Form B3.8.3

2/05/2023

TQ-BR-150-F31

(LORT)

Training Program Description: Licensed Operator

Requalification Training Program

TRM Change 23-

2

Technical Requirements Manual Change Request Form:

Appendix M

2/05/2023

25-002 (EC 645015)

SX pump Single Failure Analysis (Ref. IR 04880091)

Operability

Evaluations

EC 645071

Evaluate Past-Operability for CC HX Outlet Temperature

Limit of 120 vs 128 in 1/2BwEP ES-1.3 IR 4867564

0Bw0A ENV-3

BRAIDWOOD COOLING LAKE LOW LEVEL UNIT 0

105

0BwFSG-11

ALTERNATE SFP MAKEUP AND COOLING UNIT 0

0BwFSG-5

INITIAL ASSESSMENT AND FLEX EQUIPMENT STAGING

UNIT 0

0BwOS FX-12

FLEX PUMP FULL FLOW TEST SURVEILLANCE

0BwOS FX-1a

AAR FLEX Equipment

1Bw0A PRI-6

Component Cooling Malfunction Unit 1

110

1Bw0A PRI-8

Essential Service Water Malfunction Unit 1

107

1BwCA-0.0

LOSS OF ALL AC POWER UNIT 1

308

Procedures

1BwEP ES-1.3

Transfer to Cold Leg Recirculation - Unit 1

304 & 305

1BwEP-0

Reactor Trip or Safety Injection Unit 1

307

1BwOA ELEC-1

Loss of DC Bus - Unit 1

119

1BwOSR 3.6.6.2

Reactor Containment Fan Cooler Surveillance

1BwOSR 3.6.6.3-

SX System Flow Balance Surveillance

1BwOSR 3.6.6.3-

SX System Flow Balance Surveillance

AD-AA-101-1002

Writer's Guide for Procedures and T&RM

BD-EP ES-1.3

Transfer to Cold Leg Recirculation

301

BD-EP-0

Reactor Trip or Safety Injection

303

BwAR 1-2-A1

Low Suction Pressure Overcurrent

BwAR 1-2-A2

SX Pump DSCH HDR Press Low

BwAR 1-21-B7

Bus 141 CTL PWR Failure

BwAR 1-21-D6

25 VDC Bus 111 Ground

BwOP CC-14

Post-LOCA Alignment of the CC System

BwOP CC-8

Isolation of CC Between Units 1 and 2

CC-AA-118

DIVERSE AND FLEXIBLE COPING STRATEGIES (FLEX),

SPENT FUEL POOL INSTRUMENTATION (SFPI), AND

HARDENED CONTAINMENT VENT SYSTEM (HCVS)

PROGRAM DOCUMENT

CC-BR-118-1001

SITE IMPLEMENTATION OF DIVERSE AND FLEXIBLE

COPING STRATEGIES (FLEX) AND SPENT FUEL POOL

INSTRUMENTATION PROGRAM

CC-BR-118-1004

Braidwood Station Unit 1 & 2: Final Integrated Plan

Document Mitigation Strategies for a Beyond-Design-Basis

External Event

CC-BR-118-1005

BRAIDWOOD FLEX (BDBEE) VALIDATION PROCESS

OP-AA-103-105

LIMITORQUE MOTOR OPERATED AND CHAINWHEEL

OPERATED VALVE OPERATIONS

OP-AA-103-105

Limitorque Motor Operated and Chainwheel Operated Valve

Operators

OP-BR-102-106

OPERATOR RESPONSE TIME PROGRAM AT

BRAIDWOOD STATION

OP-BR-FX-1005

Low Head FLEX Pump Operating Guideline

01666709

1. 1 Pump Flow Water Through Pump & Re-Rack All Hoses

11/05/2014

Work Orders

01807512

MM 1SX007 Perform Mechanical and Grease Inspections

2/14/2018

01807737

1SX007 Motor Operated Valve Diagnostic Test

2/14/2018

05011128

0FX03PB Performance Test

06/26/2022

05011597

0FX02PA Performance Test

06/25/2022

05080381

1SX007 Motor Operated Valve Diagnostic Test

01/31/2024

05080382-01

MM 1SX007 Perform Mechanical and Grease Inspections

01/20/2024

05133353

IST-PIT-1SX007 U-1 SX Outlet from CC HX

09/29/2022

05144978

U1 SX System Flow Balance Surveillance

04/19/2021

05165921

Comprehensive IST Requirements for 1SX01PA

11/18/2022

270851

0FX03PA Component Operational Inspection

08/21/2023

05304779

U1 Reactor Containment Fan Cooler Surveillance

04/30/2024

05354189

0FX03PA Performance Test

06/19/2024

05605262

OP FLEX Equipment Readiness Surveillance Semiannual

06/02/2025

05605267

ASME Service Requirements for 1A Essential Service Water

Pump

2/27/2025

05627880

0FX01PB Performance Test

06/10/2025

05661943

Unit 1, 0 CPNT Cooling HTX SX WTR Availability MNTLY S

06/01/2025

1912895-02

Need Contingency WO for 1SX01PB Impeller Replacement

10/12/2016

1912896-03

2SX01PA Functional Run

11/30/2025

1916316

1SX007 MCC Thermal

O.L. Prot Srv, E1AP21E, MCCA, T40

2/14/2018

4883959-02

2B SX Pump PMT

04/30/2020

213718-01

1SX01PA Bus 141 Cub 02 1A SX Pump Relay Routine

04/19/2024

284482-01

1PA23J SAT 142-1 Protective Relay Routine

05/12/2025

24792-01

D/G 1APH A OC

06/22/2025