ML25059A249
| ML25059A249 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 03/14/2025 |
| From: | Kerri Kavanagh NRC/NRR/DRO/IQVB |
| To: | Fosaaen C NuScale |
| References | |
| EPID I-2025-201-0004 IR 2025001 | |
| Download: ML25059A249 (1) | |
Text
Ms. Carrie Fosaaen Vice President, Regulatory Affairs NuScale Power LLC 1100 NE Circle Blvd., Suite 200 Corvallis, OR 97330
SUBJECT:
NUSCALE POWER, LLC., STANDARD DESIGN APPROVAL APPLICATION AIRCRAFT IMPACT ASSESSMENT INSPECTION, NUCLEAR REGULATORY COMMISSION INSPECTION REPORT NO. 05200050/2025-201
Dear Ms. Fosaaen,
From February 10, 2025, through February 14, 2025, the U.S. Nuclear Regulatory Commission (NRC) conducted an inspection of the NuScale Power LLC., (NuScale) Standard Design Approval Application Aircraft Impact Assessment (AIA). The NRC staff performed this inspection at the NuScale Corporate office located in Corvallis, Oregon. The purpose of the inspection was to assess NuScales compliance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.150, Aircraft impact assessment. The enclosed report presents the results of this inspection.
Based on the limited sample of documents reviewed, no violations were identified, and the NRC inspection team concluded the NuScale is in compliance with the requirements of 10 CFR 50.150.
In accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure(s), and your response will be made available electronically for public inspection in the NRCs Public Document Room or from the NRCs document system ADAMS, accessible at http://www.nrc.gov/readingrm/adams.html.
March 14, 2025
C.Fosaaen 2
Sincerely, Kerri A. Kavanagh, Chief Quality Assurance Vendor Inspection Branch Division Inspection Programs and Regional Support Office of Nuclear Reactor Regulation Docket No.: 05200050 EPID No. I-2025-201-0004
Enclosures:
1.
Inspection Report No. 05200050/2025-201 2.
Attachment Signed by Kavanagh, Kerri on 03/14/25
ML25059A249 NRR-106 OFFICE NRR/DRO/IQVB NRR/DSS/SNRB NRR/DEX/ESEB NAME Frankie Vega RNolan GWang DATE 3/4/2025 3/4/2025 3/13/2025 OFFICE NRR/DRA/APLB NRR/DNRL/NRLB NRR/DRO/IQVB NAME TDinh SJoseph KKavanagh DATE 3/4/2025 3/4/2025 3/14/2025
Enclosure U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION DIVISION OF REACTOR OVERSIGHT STANDARD DESIGN APPROVAL APPLICATION AIRCRAFT IMPACT ASSESSMENT INSPECTION REPORT Docket No.:
05200050 Report No.:
05200050/2025-201 Applicant: NuScale Power, LLC 1100 NE Circle Boulevard, Suite 200 Corvallis, OR 97330 Applicant
Contact:
Ms. Carrie Fosaaen Senior Director, Regulatory Affairs NuScale Power, LLC Email: cfosaaen@nuscalepower.com Office: (541) 452-7126 Inspection Location:
1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Nuclear Industry Activities:
NuScale Power LLC. has completed its aircraft impact assessment of the standard design approval application to comply with the U.S. Nuclear Regulatory Commission requirements in Title 10 of the Code of Federal Regulations Section 50.150, Aircraft impact assessment.
Inspection Dates:
February 10-14, 2025 Inspectors:
Frankie Vega, NRR/DRO/IQVB Ata Istar, NRR/DEX/ESEB Ryan Nolan, NRR/DSS/SNRB George Wang, NRR/DEX/ESEB Thinh Dinh, NRR/DRA/APLB Stacy Joseph, NRR/DNRL/NRLB Approved by:
Kerri A. Kavanagh, Chief Quality Assurance Vendor Inspection Branch Division Inspection Programs and Regional Support Office of Nuclear Reactor Regulation
ATTACHMENT Att1-1 EXECUTIVE
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) conducted this inspection to verify that NuScale Power LLC., (NuScale) had implemented the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.150, Aircraft impact assessment, (10 CFR 50.150) and performed a design-specific assessment of the effects on the facility of the impact of a large commercial aircraft.
The NRC conducted the inspection at the NuScale Corporate office in Corvallis, Oregon on February 10-14, 2025.
The following served as the basis for the NRC inspection:
10 CFR 50.150 During this inspection, the NRC inspection team implemented Inspection Procedure (IP) 37804, Aircraft Impact Assessment, dated March 23, 2020.
This inspection was performed to verify that NuScales aircraft impact assessment (AIA) of the NuScale Power Plant US460 standard design complies with the requirements of 10 CFR 50.150. The Nuclear Energy Institute (NEI) 07-13, Methodology for Performing Aircraft Impact Assessments for New Plant Designs, Revision 8, dated April 2011, has been endorsed by the NRC in Regulatory Guide (RG) 1.217, Guidance for the Assessment of Beyond-Design-Basis Aircraft Impacts, as one means of performing an AIA acceptable to the NRC. NuScale utilized NEI 07-13, Revision 8, with no exceptions, to perform its AIA.
For the implementation of this inspection, the NRC inspection team used Revision 1 of NuScales standard design approval application (SDAA). The results of the inspection are summarized below.
Systems-Loss Assessment The NRC inspection team concluded that the systems-loss assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
Fire Damage Assessment The NRC inspection team concluded the fire damage assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
Structural Damage Assessment The NRC inspection team found that the structural damage assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
Documentation and Quality Assessment The NRC inspection team concluded that the documentation and quality assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
ATTACHMENT Att1-2 REPORT DETAILS 1.
Systems-Loss Assessment
- a. Inspection Scope The NRC inspection team verified that the aircraft impact assessment (AIA) adequately addressed a system-loss assessment consistent with the requirements of 10 CFR 50.150. Specifically, the NRC inspection team determined that the systems-loss assessment included:
verification of the location of those structures, systems, and components (SSCs) that provide core cooling or containment isolation, and spent fuel pool (SFP) integrity to determine the potential for damage by aircraft impact, verification that those SSCs would be capable of performing their intended function given the established structural, shock, and fire damage footprints and the rule sets and assumptions provided in Nuclear Energy Institute (NEI) 07-13, Methodology for Performing Aircraft Impact Assessments for New Plant Designs, Revision 8, dated April 2011 (NEI 07-13),
verification that NuScale addressed accident initiators, such as a breach of the reactor coolant system (RCS) or the failure of the reactor to trip, that could result from damage caused by an aircraft impact; and verification that success paths for core cooling exist.
a.1 Determination of the location of credited SSCs The NRC inspection team reviewed NuScales selection of SSCs needed to prevent fuel damage in the core and the documented spatial configuration of those SSCs.
NuScale indicated that its objective in adding key design features to address the AIA rule was to maintain core cooling, intact containment, and SFP integrity. Therefore, SSCs needed to provide for active SFP cooling were not reviewed by the NRC inspection team.
The NRC inspection team compared the descriptions of SSCs in the assessment to those in the Final Safety Analysis Report (FSAR) and the probabilistic risk analysis (PRA) and reviewed whether the scope of SSCs treated in the assessment was complete and consistent with those needed to satisfy the core cooling success criteria in the PRA. The NRC inspection team used equipment location data and drawings from the general arrangement drawings to confirm that the locations of equipment documented in the assessment were accurate.
a.2 Determination of the state of SSCs in the aircraft impact scenarios The NRC inspection team reviewed the AIA to determine whether NuScale had correctly applied the rules and assumptions given in NEI 07-13 for the loss of SSCs.
Specifically, the NRC inspection team reviewed the SSCs that NuScale had identified in the AIA as remaining functional in each scenario and verified that the
ATTACHMENT Att1-3 basis used to conclude that these SSCs will survive the conditions created by an aircraft impact is consistent with the rules and assumptions given in NEI 07-13. For example, for each at power strike scenario, the NRC inspection team verified that a decay heat removal pathway exists using the decay heat removal system (DHRS) and emergency core cooling system (ECCS). For all strike scenarios, the NRC inspection team verified that necessary support SSCs, such as containment isolation valves and the ultimate heat sink, were available. The NRC inspection team determined that the potential effect of structural, shock, or fire damage will not prevent the core cooling equipment and the credited systems from remaining capable of performing their intended function following an aircraft impact.
a.3 Determination of accident conditions The NRC inspection team verified that NuScale used appropriate assumptions and scenarios to determine accident conditions. These assumptions were consistent with NEI 07-13 and include:
NuScales success criteria and the scenario analysis that addresses initial plant states of 100 percent power and cold shutdown, the analysis takes no credit for the availability of offsite power, the consideration of the possibility of an anticipated transient without scram (ATWS), and NuScale has considered the influence of containment status on the operability of other equipment.
Specifically, the NRC inspection team reviewed NuScales treatment of the following potential accident conditions:
Loss of Coolant Accident (LOCA) inside containment The NRC inspection team reviewed NuScales assessment of a LOCA inside the containment to determine if the containment is adequately protected such that it could not be impacted by an aircraft. The NRC inspection team determined that the assessment adequately demonstrated that neither shock damage nor physical damage to the containment vessel would occur and, as such, verified that a LOCA inside containment would not occur.
LOCA outside containment The NRC inspection team reviewed NuScales assessment of a LOCA outside containment to determine if piping outside of primary containment that is connected to the reactor coolant pressure boundary, above grade level, is protected from structural damage. The NRC inspection team used plan and elevation drawings of the reactor building (RXB) to confirm that NuScales assessment effectively determined that the applicable piping and isolation valves are adequately protected from structural damage.
ATWS The NRC inspection team reviewed the AIA to determine if NuScale adequately assessed the potential for any damage scenarios that could affect the ability to trip
ATTACHMENT Att1-4 the reactor. The NRC inspection team verified that an ATWS was not a viable outcome from an aircraft impact because the equipment necessary to maintain the reactor shutdown is outside the damage footprint areas or a loss of power to the equipment will induce a reactor trip.
Flooding The NRC inspection team reviewed the AIA to determine if NuScale adequately assessed the potential for internal and external flooding from large water sources as described in NEI 07-13. The assessment stated that there are no pipe breaks which would cause flooding of the RXB from an aircraft impact that is more severe than existing pipe break flooding analysis. The NRC inspection team verified that any potential large water source was either not vulnerable, was bounded by the internal flooding analysis, or did not adversely impact credited fuel cooling equipment.
Loss of Decay Heat Removal-Shutdown The NRC inspection team reviewed the AIA to determine if NuScale adequately assessed the potential for a loss of decay heat removal event when the reactor is shutdown. The NRC inspection team reviewed the assumptions used in the analysis and verified that applicable assumptions were consistent with guidance in NEI 07-13.
During refueling the lower portion of the containment and reactor vessels will be completely submerged in the protected RXB pool. Therefore, core cooling is maintained during shutdown conditions.
b.4 Identification of Success Paths The NRC inspection team reviewed the AIA to determine if NuScale had a success path for core cooling. The NRC inspection team reviewed the information documented in Chapter 19 of the FSAR, Revision 1, and supporting PRA documents such as system dependencies matrix and accident sequence notebook. The NRC inspection team verified that the core cooling methods identified by NuScale are shown as success paths for avoiding core damage.
The NRC inspection team also discussed the systems-loss assessment with NuScales management and technical staff. The attachment to this inspection report lists the documents reviewed and the staff interviewed by the NRC inspection team.
- b. Observations and Findings No findings of significance were identified.
- c. Conclusions The NRC inspection team concluded that the system-loss assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
2.
Fire Damage Assessment a.
Inspection Scope The NRC inspection team verified that the AIA adequately assesses fire damage consistent with the requirements of 10 CFR 50.150 to ensure that plant SSCs maintain
ATTACHMENT Att1-5 key safety functions (core cooling, containment, SFP cooling, and SFP integrity).
Specifically, the NRC inspection team verified that the fire damage assessment included:
the AIA adheres to applicable assumptions and methodologies prescribed in NEI 07-13 for performing AIA, identification and incorporation of the necessary design features and functional capabilities, a realistic and design-specific assessment, a description of key design features credited in the AIA that are consistent with those documented in the FSAR, an assessment that damage footprints include the effects from the spread of fire damage through existing connected compartments and through new compartment connections due to overpressure, a verification that the SSCs credited for safe shutdown following aircraft impact scenarios remain free from physical and fire damage, and a verification that the SFP integrity is maintained.
a.1 Fire Damage Footprint Assessment The NRC inspection team verified NuScales AIA methodologies that led NuScale to conclude that NuScales credited structural design features prevented damages to the core cooling equipment and maintained SFP integrity. Specifically, the NRC inspection team verified that there was no fire damage caused by aircraft impact in the vicinity of essential SSCs needed to maintain reactor core and SFP cooling. The NRC inspection team verified that the AIA followed NEI 07-13, Section 3.3, Step 3: Damage Footprint Assessment, guidance in assessing the damage footprint due to debris, shock and fire spread from an aircraft impact. The NRC inspection team verified that the AIAs postulated strike locations covered all elevations and building faces that are susceptible to an aircraft strike. The NRC inspection team also verified consistency between the FSAR and AIA to assure plant fire areas and design features credited in the AIA are as described in the FSAR.
In addition, the NRC inspection team assessed NuScales preventative measures and credited protections to exterior openings. Specifically, the NRC inspection team verified that NuScales assessment of credited structural design features, including 3-hour rated, 5 pounds per square inch differential (PSID) steel-composite fire barriers, the RXB equipment door and the heating, ventilation and air conditioning awning located at the perimeter of the RXB, adequately prevented physical damages from propagating to the interior regions.
a.2 Fire Damage Effects on SSCs The NRC inspection team reviewed the AIA to determine if NuScale assessed the fire damage effects on SSCs; however, since there was no fire damage to areas where
ATTACHMENT Att1-6 essential SSCs needed to maintain reactor core cooling are located, this assessment is not required.
The NRC inspection team also discussed the fire damage assessment with NuScales management and technical staff. The attachment to this inspection report lists the documents reviewed and the staff interviewed by the NRC inspection team.
b.
Observations and Findings No findings of significance were identified.
- c. Conclusions The NRC inspection team concluded that the fire damage assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
3.
Structural Damage Assessment a.
Inspection Scope The NRC inspection team verified that the AIA was sufficiently rigorous and realistically evaluated a design-specific structural damage analysis of the effects of the impact of a large, commercial aircraft on the facility consistent with the requirements of 10 CFR 50.150: Specifically, the NRC inspection team verified that the structural damage assessment included:
adequate information found in plant documentation including plant arrangement drawings displaying locations of major equipment and plant elevation drawings documenting the relative heights of the RXB, civil-structural drawings that provide wall thicknesses, reinforcement details, steel-plate composite (SC) wall details, and material specifications, general structural analysis considerations such as design inputs, analysis parameters and assumptions, computer codes, methods used for structural analyses and results to evaluate whether NuScale adequately analyzed the effects of, and damage to structures resulting from, global and local aircraft impact loads, analytical evaluation and experimental verification of the RXB external SC walls subjected to the aircraft impact loading based on recommendations set forth in Section 2.4.1 (4) of NEI 07-13, SFP aircraft impact analyses to evaluate whether NuScale addressed the criteria in section 2.5.2 of NEI 07-13, and structural damage footprint assessments to evaluate whether NuScale adequately assessed the RXB containing essential SSCs for maintaining reactor core cooling using the damage rule sets in Section 3.3 of NEI 07-13.
ATTACHMENT Att1-7 a.1 Structural Assessment Document Review The NRC inspection team reviewed NuScales structural assessment design inputs, including plant arrangement drawings, plant elevation drawings, civil-structural drawings, and material specifications. The NRC inspection team verified that the plant arrangement drawings display the locations of major equipment (e.g., RXB crane), and the plant elevation drawings identified the relative heights of the RXB.
a.2 General Structural Analysis NuScales AIA evaluated 13 structural analysis cases for the assessment of the RXB including connections between SC walls and roofs, and 4 structural analysis cases for the assessment of crane stability in accordance with NEI 07-13. The analyses performed for the above ground portion of the RXB confirmed that the RXB pool liner would remain intact.
The NRC inspection team verified that NuScale used appropriate design inputs including the structural analysis parameters and assumptions, type of finite elements used in each analysis, material models considered, boundary conditions and extent of model, initial conditions, and time duration of the analysis. The NRC inspection team verified that NuScale adequately documented and justified the structural design input for a sampling of analyses and adequately analyzed the effects of, and damage to structures resulting from, local and global loading arising from an aircraft impact.
In addition, the NRC inspection team verified that NuScale properly modeled the reinforcing bars as sub-elements embedded within the concrete elements at the appropriate locations, and the SC walls using the elements for concrete, steel plates, shear tie plates, vertical rib plates, and studs.
The NRC inspection team verified that all potential aircraft impact scenarios were considered in the structural analyses. The NRC inspection team reviewed a sample of the structural damage impact scenario analyses and verified that NuScale properly applied the NRC-supplied forcing function and missile-target interaction to the appropriate structural damage impact scenarios. In addition, the NRC inspection team reviewed the assumptions used in the structural damage analyses and verified that NuScale adequately documented the technical basis and assumptions used in the analyses.
The NRC inspection team reviewed a sample of structural damage analyses and verified that NuScale used the correct failure criteria. As part of the review, the NRC inspection team reviewed the various material properties used in the structural analyses, including concrete strength of 5,000 psi (34.5 MPa) for the SC walls and 7,000 psi (48.3 MPa) used for the slabs at various elevations of the RXB. In addition, the NRC inspection team confirmed that the analyses used the Dynamic Increase Factor (DIF) based on NEI 07-13.
a.3 RXB Specific Impact Assessment The NRC inspection team reviewed the RXB impact analyses to evaluate whether NuScale met the sufficiency criteria in NEI 07-13, Section 2.5.
ATTACHMENT Att1-8 The NRC inspection team reviewed the structural damage assessment as it relates to local loading on the RXB structure and verified that NuScale conducted the following activities in accordance with NEI 07-13, Section 2.1:
documented and cross-checked the aircraft engine parameters used in the analysis against NRC-specified parameters, properly applied the various local loading formulas referenced in Design of Composite SC Walls to Prevent Perforation from Missile Impact, International Journal of Impact Engineering 75 (2015) 75-87 and NEI 07-13, Subsection 2.1.2, to arrive at the degree of local damage and the wall thickness required to prevent perforation of the target, and used the formulas cited in NEI 07-13.
The NRC inspection team reviewed the structural damage assessment as it relates to global loading effects on the RXB structure. The NRC inspection team verified that the following activities were conducted in accordance with NEI 07-13, Section 2.2:
documentation and use of the application of the force time-history analysis method and cross-checking it for its equivalency to the NRC-specified force time-history, documentation of the application of the missile-target interaction analysis method and cross-checking it for its equivalency to the NRC-specified force-time history, the missile-target interaction analysis method reasonably captured the mass distribution of the missile to determine the missile-target interaction from the force-time history, and for the application of the force time-history analysis method, NuScale properly used and adequately documented the NRC-specified spatial distribution of the impact force in the analyses.
The NRC inspection team reviewed a sample of documents for material characterization and failure criteria related to the structural damage assessment and verified that the following analysis activities were conducted in accordance with NEI 07-13, Section 2.3:
application of the LS-DYNA concrete constitutive model consisting of material properties and the equations used to model the nonlinear behavior of both steel and reinforced concrete materials in the analyses. The model parameters used are adequately documented and consistent with the material properties and equations documented in NEI 07-13, Section 2.3, application of the ductile failure strain limits specified in NEI 07-13, Subsection 2.3.2, for the various materials used in the analyses,
ATTACHMENT Att1-9 the concrete structural failure criteria used in the analyses are appropriately documented and consistent with the criteria specified in NEI 07-13, Subsection 2.3.3, application and documentation of the material models specified in NEI 07-13, Subsection 2.3.4, and application and documentation of the structural integrity failure criteria specified in NEI 07-13, Subsection 2.3.5.
The NRC inspection team reviewed the major assumptions applied to the containment and SFP related structural analyses and verified that the following activities were conducted in accordance with NEI 07-13, Section 2.4:
missile-target interaction analysis model properly assumed that the aircraft impact was perpendicular to the centerline of the containment, missile-target interaction analysis model properly assumed takeoff weight such that the missile-target interaction model is equivalent to the NRC-specified force time-history, containment regions containing critical penetrations received an appropriate level of special consideration, RXB aircraft impact analyses properly assumed that both the engine and the aircraft fuselage strike were perpendicular to and at the mid-point of the RXB
- wall, assessment of potential aircraft impact at other locations that could result in greater consequences, potential aircraft impacts on the RXB and its effects on its pool water inventory were properly considered, and potential aircraft impacts on the exterior walls of the RXB and its potential effects on the RXB crane falling into the RXB pool and damaging the NuScale Power Modules were properly considered.
The NRC inspection team reviewed the RXB including SFP related structural analyses and verified NuScales conclusion that the damage imparted to the RXB walls would not result in leakage of the RXB pool water. Thus, the integrity of the RXB of pool water is maintained, consistent with the sufficiency criteria of NEI 07-13, Subsection 2.5.2.
a.4 Structural Damage Footprint Assessment The NuScale AIA evaluated a total of 17 different impact scenarios in accordance with NEI 07-13. The NRC inspection team reviewed all the impact scenarios associated with structural performance.
ATTACHMENT Att1-10 The NRC inspection team reviewed the structural damage footprint analyses to evaluate whether the following items of interest related to the damage rule sets identified in NEI 07-13, Section 3, "Heat Removal Capability," have been met. These items of interest include:
structures of concern that contain SSCs have been identified, a systematic evaluation of susceptible damage was conducted and adequately documented, and assumptions used to determine elevations of concern have been addressed and adequately documented.
The NRC inspection team verified that the structural damage rule sets for RXB and including SFP structures were appropriately assessed consistent with the guidance in NEI 07-13, Subsection 3.3.1. The NRC inspection team verified that the following activities were conducted in the analyses:
various impact points have been investigated and documented in order to define the damage footprint, structural damage rule sets regarding perforations were appropriately developed, shock damage was evaluated in the structural damage footprints and these evaluations have been adequately documented, the guidance in NEI 07-13, Table 3-3, was used to define the shock damage footprints and was adequately documented, and shock effects impacting seismic separation between buildings has been adequately assessed and documented.
The NRC inspection team also discussed the structural damage assessment with NuScales technical staff and its consultants. The attachment to this inspection report lists the documents reviewed and the staff interviewed by the NRC inspection team.
b.
Observations and Findings No findings of significance were identified.
c.
Conclusions The NRC inspection team concluded that the structural damage assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
4.
AIA Documentation and Quality Assessment a.
Inspection Scope
ATTACHMENT Att1-11 The NRC inspection team verified that the NuScale performed a quality assessment to ensure the AIA was documented and maintained consistent with the requirements of 10 CFR 50.150. The NRC inspection team confirmed that:
NuScale adequately documented the quality assessment consistent with NEI 07-13, Section 5.1; and NuScale adequately established standards and measures to establish the validity of the assessment and supporting calculations.
The NRC inspection team reviewed EP-0303-52592, Engineering Change Control, dated September 16, 2022, which provide the administrative controls governing the design change process and reviewed a sample of completed design change requests to verify that any effects of design changes on key design features credited in the AIA were adequately identified and evaluated.
The NRC inspection team reviewed the supplier evaluation review forms used by NuScale to review and approve suppliers design deliverables. Specifically, the NRC inspection team reviewed supplier deliverable review forms associated with the AIA structural and heat removal assessments and determined the forms were completed in accordance with procedures and issues identified were adequately dispositioned.
The NRC inspection team reviewed the LS-Dyna software benchmarking analysis report which established a modeling methodology for dynamic impact problems including defining the material model, contacts, constraint equations, and boundary conditions used for the AIA structural analysis. The NRC inspection team determined that NuScale had adequately documented the validation and benchmarking activities associated with the use of LS-Dyna for the AIA, consistent with Appendix C of NEI 07-13.
The NRC inspection team also discussed the quality assessment with NuScales technical staff and its consultants. The attachment to this inspection report lists the documents reviewed and the staff interviewed by the NRC inspection team.
b.
Observations and Findings No findings of significance were identified.
c.
Conclusions The NRC inspection team concluded that, the documentation and quality assessment performed by NuScale for the AIA is consistent with the regulatory requirements of 10 CFR 50.150.
5.
Entrance and Exit Meetings On February 10, 2025, the NRC inspection team discussed the scope of the inspection with Ms. Carrie Fosaaen, NuScale Vice President, Regulatory Affairs and other representatives from NuScale. On February 13, 2025, the NRC inspection team presented the inspection results and observations during an exit meeting with Ms. Carrie Fosaaen, NuScale VP, Regulatory Affairs and representatives from NuScale.
ATTACHMENT Att1-12 1.
Person Contacted Name Title Affiliation Entrance Exit Interviewed Mark Shaver Director, Regulatory Affairs NuScale X
X Kathy Warnock Supervisor, Quality Assurance Oversight NuScale X
X X
Elisa Fairbanks Supervisor, Licensing NuScale X
X X
Ross Snuggerud Chief Engineer -
Operations NuScale X
X J.J. Arthur Vice President, Engineering NuScale X
X Stephanie Terwilliger Program Manager, Licensing NuScale X
X X
Jess Sloan Licensing Coordinator NuScale X
X X
Randy James Principal SSC, LLC X
X X
Dan Parker Senior Consultant Structural Integrity Associates X
X X
Peter Shaw Licensing Engineer NuScale X
X X
Gary Hayner Director Jensen Hughes X
X X
Daniel Larson Manager Jensen Hughes X
X X
Rim Nayal Engineering Manager NuScale X
X X
Josh Parker Senior Manager, Plant Engineering NuScale X
X Carrie Fosaaen Vice President, Regulatory Affairs NuScale X
X Erwin Laureano Manager, Services NuScale X
X Doug Bowman Manager, Services NuScale David Benson Civil/Structural Engineer 4 NuScale Jeff Ehlers Senior Manager, Plant Systems Engineering NuScale Robert Gamble Senior Vice President, Engineering NuScale Sangeet Gupta Contractor, Mechanical Systems Engineering/Fire Protection X
ATTACHMENT Att1-13
- Remote 2.
Inspection Procedures Used Inspection Procedure 37804, Aircraft Impact Assessment, dated March 23, 2020 3.
Documents Reviewed AIA Reports ER-124770, Aircraft Impact Assessment for NuScale NP-12 Plant Design (Applied to US 460), Revision 0 ER-124795, Aircraft Impact Assessment for NuScale US460 Reactor Building, Revision 2, dated January 28, 2025 ER-113229, NuScale Design US 460 AIA Heat Removal Report, Revision 2, dated February 6, 2025 ER-124779, Aircraft Impact Assessment for NuScale NP-12 Plant Design, Revision 0.
Sean Park Civil-Structural Engineer NuScale X
X Sarah Bristol Manager, Probabilistic Risk Assessment NuScale X
X Thomas Griffith Manager, Licensing NuScale X
Gary Becker Senior Regulatory Affairs Counsel NuScale X
Rose Charoensombud Manager, Mechanical Systems NuScale X
Frankie Vega Inspector, Team Lead NRC X
X Stacy Joseph Senior Project Manager NRC X
X Ata Istar Civil Engineer (Structural)
NRC X
X Ryan Nolan Senior Nuclear Engineer NRC X
X George Wang Civil Engineer (Structural)
NRC X
X Thinh Dinh Fire Protection Engineer NRC X
X Kerri Kavanagh Branch Chief NRC X*
X Prosanta Chowdhury Senior Project Manager NRC X*
Mahmoud Jardaneh Branch Chief NRC X*
Getachew Tesfaye Senior Project Manager NRC X*
X*
ATTACHMENT Att1-14 Systems-Loss Assessment SD-116186, DHRS System Design Description, Revision 5, dated June 3, 2022 ER-102579, Failure Modes and Effects Analysis for the Emergency Core Cooling System, Revision 1, dated September 21, 2022 ER-102067, Accident Sequence Analysis Notebook, Revision 1, dated December 21, 2022 ER-115691, US460 Safe Shutdown Systems Determination, Revision 0, dated June 21, 2022 ER-102068, Success Criteria Notebook, Revision 1, dated December 20, 2022 Fire Damage Assessment EQ-100845, Penetration Seals, Revision 0, dated March 20, 2021 EQ-126998, Reactor Building Equipment Door Specification, Revision 2, dated January 17, 2025 EC-156190, Reactor Building Local Design Details for Aircraft Impact Assessment, Revision 1 ED-109868, Steel Composite Modular Assembly Typical Details and General Notes, Revision 3, dated March 23, 2023 ER-120866, Fire Hazards Analysis, Revision 0, dated November 22, 2022 ER-120707, Post-Fire Safe Shutdown Analysis, Revision 0, dated November 15, 2022 NuScale US 460 SDAA Final Safety Analysis Report, Chapter 1.2, General Plant Description, Revision 2 NuScale US460 SDAA Final Safety Analysis Report, Chapter 9A, Fire Hazards Analysis, Revision 2 NuScale US460 SDAA Final Safety Analysis Report, Chapter 19.5, Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts, Revision 2 Structural Damage Assessment EC-114628, Reactor Building Structural Design, Revision 5 ED-109868, Steel Composite Modular Assembly Typical Details and General Notes, Revision 3, dated March 20, 2023 ED-118614, Structural Steel General Notes and Details, Revision 0, dated April 6, 2022 EC-128707, Reactor Building Structural Steel Calculation, Revision 2, dated May 21, 2024 EC-118854, Structural Qualification of NuScale 950 Ton Reactor Building Crane, Revision 1, dated May 26, 2023 ER-178981, LS DYNA Benchmarking for Impact Problems, Revision 0, dated January 28, 2025 Engineering Change Reports (ECRs)
ECR 177887 Additional AIA barriers for RXB, dated February 3, 2025 ECR 177888 Increase FP thickness for RBED, dated December 17, 2024
ATTACHMENT Att1-15 ECR 178686 Re-Locate the OPVE pipes, dated January 17, 2025 ECR 178879 Revision to EC 128707 to modify slab support requirements, dated January 27, 2024 ECR 177100 Increase size of vestibule (room 510) located on east side of RXB El. 100, dated December 9, 2024 Drawings ED-111239, Reactor Building General Arrangement Drawings, Revision 1, dated December 18, 2024 ED-119156, Reactor Building Fire Area Plan View Drawings, Revision 1, dated December 18, 2024 ED-103683, NuScale 6 Module Plant Overall Plot Plan, Revision 1, January 23, 2023 ED-128718, Reactor Building Steel-Composite Wall Elevation Drawings, Revision 1, dated February 5, 2025 ED-122956, Concrete Drawings for the Reactor Building, Revision 1, dated February 5, 2025 ED-122758, Reactor Building SDA Critical Sections, Revision 2, dated December 9, 2022 ED-129355, RXB (F010) Structural Steel Drawings, Revision 0, dated October 26, 2022 ED-130795, Reactor Building Crane Mechanical/Structural Drawings, Revision 1, dated March 13, 2023 ED-101258, Concrete General Notes (GN) for SC-II and SC-III Applications, Revision 1, dated October 3, 2022 Quality Assessment Engineering Procedure (EP) EP-0303-52592, Engineering Change Control, Revision 12, dated September 16, 2022 EP-0303-3340, Preparation and Approval of Engineering Documents, Revision 15, dated January 24, 2023 EP-0303-309, Preparation and Approval of System Design Descriptions, Revision 13, January 24, 2023 EP-0303-312, Preparation and Approval of Engineering Specifications, Revision 24, dated November 20, 2024 EP-0303-315, Preparation and Approval of Engineering Drawings, Revision 16, dated May 18, 2023 EP-0303-303, Preparation and Approval of Engineering Calculations, Revision 19, dated January 21, 2023 QP-0303-10267, Design Control Process, Revision 22, dated July 31, 2024 EP 0703-1417, Supplier Deliverable Review and Approval (Design and Testing),
Revision 19, dated June 18, 2024.
SW-175083 Aircraft Impact Assessment Inspection Support-Heat Removal, Revision 0, Revision 0, dated October 14, 2024 SW 175082 Aircraft Impact Assessment Inspection Support-Structural, Revision 2, dated January 20, 2025 TO #24 for SW-175082 Revision 1, dated December 17, 2024 TO #8 for SW-175083, Revision 0, dated December 23, 2024
ATTACHMENT Att1-16 ER 178981 LS-Dyna Benchmarking for Impact Problems, Revision 0, dated August 2022.
Supplier deliverable Review Form for Aircraft Impact Assessment for NuScale, dated 02/06/2025 Supplier Deliverable Review Form for NuScale Design US460 AIA Heat Removal Report, dated 02/04/2025 Supplier Deliverable Review Form for ER-178981, dated January 28, 2025 Supplier Deliverable Review Form for ER-124797, dated February 13, 2025 Condition Report CR 176293 A requirement related to the limitation of fire propagation under AI is not implemented, dated November 29, 2024.