ML25293A457

3 - Safety-Significant SSC Design Requirements

ML25293A457
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	10/20/2025
From:	Matthew Hiser, Bill Lin, Hanh Phan, Radel T Advisory Committee on Reactor Safeguards, NRC/NRR/DANU/UAL1, NRC/NRR/DANU/UTB2
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Download: ML25293A457 (1)

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NRC Review of Kemmerer Unit 1 Construction Permit Application Safety-Significant SSC Design Requirements Hanh Phan, Matt Hiser, Bruce Lin, Tracy Radel NRR/DANU/UTB2 and UAL1 ACRS Subcommittee Meeting October 21-23, 2025

Topics

• Design Basis Hazard Levels (DBHLs)

• Seismic Classification and Design Criteria

• Special Treatments for Safety-Significant SSCs

• Programmatic Special Treatments 2

Regulatory Requirements and Guidance Applicable Regulations

• 10 CFR 50.34(a), 50.35, 50.150(a)(1): Site characterization, safety analysis, and aircraft impact protection

• 10 CFR Part 50, Appendix S: Seismic and fire protection requirements for nuclear power plants

• 10 CFR 100.23: Natural phenomena design criteria Guidance for Review

• DANU-ISG-2022-01: Interim staff guidance for hazard assessment

• RG 1.76, 1.78, 1.91, 1.189, and 1.204: Provide methodologies for tornadoes, chemical hazards, explosions, fire protection, and lightning protection

• RG 1.233 and RG 1.253: Technology-inclusive, risk-informed, performance-based guidance for non-light-water reactor licensing; endorses NEI 18-04 and NEI 21-07

• NUREG-0800, Sections 3.2-3.7: Standard Review Plan for the review of internal and external hazards 3

Relevant Principal Design Criteria (PDCs)

• PDC 2 - Protection Against Natural Phenomena: Ensures SSCs can withstand hazards such as wind, tornadoes, seismic events, floods, and extreme weather

• PDC 4 - Environmental and Dynamic Effects: Addresses effects from dynamic loads, fluid discharges, and missiles resulting from equipment failures

• PDC 81 - Reactor Building Design Basis: Ensures reactor building structures provide required safety margins and house safety-significant SSCs 4

DBHLs Review Approach Verify that the established DBHLs reasonably encompass all applicable internal and external hazards Review the characterization of internal and external hazards and the design strategies to mitigate their effects on SSCs Confirm that relevant hazards are comprehensively evaluated using site-specific data Examine the specific parameters for internal and external hazards Ensure no adverse impact on SSC safety functions 5

High Wind and Tornado Hazards High Winds

• Design wind speeds: 110-115 mph (3-second gusts), in accordance with ASCE/SEI 7-16

• Based on local historical data for the KU1 site (PSAR Section 2.4.1.3.2, Table 2.1-1)

• Structural design ensures SSCs maintain required safety functions under high wind conditions

• High wind hazard reasonably characterized Tornado

• Consistent with RG 1.76, Rev. 1, and its tornado parameters: maximum wind speed, translational/rotational speed, radius of maximum rotation, and pressure drop

• Site-specific data used to define design-basis tornado parameters (PSAR Section 2.4.1.3.3, Table 2.1-1)

• SSCs designed to resist tornado forces, wind-generated missiles, and pressure variations

• Tornado hazard reasonably characterized 6

Flood Hazards External Flooding

• Design-basis scenarios include river/stream flooding, dam breaches, surges, seiches, tsunamis, snow melts/ice jams, and channel diversion

• Estimated flood data for the KU1 site provided in PSAR Section 2.5.1 and Table 2.1-1

• SR SSCs located above design-basis flood levels

• Barriers and elevated structures provided for NSRST and NST SSCs

• External flood hazard reasonably characterized Internal Flooding Analysis considers worst-case postulated system failures involving sodium, molten salt, water, or oil Staff audited supporting documentation confirming PSAR Section 6.1.1.3 information Mitigation features include isolation valves, physical barriers, drains, and separation of energy island systems from the nuclear island Sodium, salt, and water leaks evaluated considering their chemical/reactive properties in design Internal flood hazard reasonably characterized 7

Missile Hazards Turbine Missiles

• Turbine generator located ~500 feet south of the nuclear island, rotor oriented north-south

• Rotor orientation and distance provide sufficient separation to protect SR SSCs from potential missile impacts

• Turbine missile hazard reasonably characterized Generated Missiles

• Hazards considered: pressurized systems, rotating equipment, secondary missiles, wind/tornado, and site proximity

• Design basis missiles from pressurized components established based on probability of adverse impact on SR SSCs of >1x10/year

• Mitigation includes physical separation, protective barriers, or SSC design to withstand a single missile impact

• Internally generated missile hazard reasonably characterized 8

Pipe Ruptures and Seismic Hazards Pipe Ruptures

• High-energy fluid piping not located in SR SSC areas; moderate-energy pipe ruptures postulated

• Mitigation includes physical separation, guard piping, fluid collection systems, or demonstration of no adverse impact

• Pipe rupture hazard reasonably characterized Seismic

• Site-specific seismic hazard characterized in PSAR Sections 2.6 and 6.4.1

• SSCs designed to withstand postulated seismic events with adequate design margin

• Seismic hazard reasonably characterized 9

Extreme Winter Precipitation and Lightning Hazards Extreme Winter Precipitation

• Loads from snow, ice, and extreme winter weather included in SSC design

• Parameters and bases described in PSAR Section 2.4.1.3.5 and Table 2.1-1

• Enclosed SSCs protected by buildings; exposed SSCs designed per safety/risk significance

• Mitigation provides confidence in maintaining safety functions under extreme winter conditions

• Extreme winter precipitation hazard reasonably characterized Lightning

• Grounding, earthing, and lightning protection systems protect personnel and equipment (PSAR Section 1.1.4.3.11)

• Nuclear island buildings and circuits protected against transient over-voltages

• Thunderstorm/lightning hazard reasonably characterized 10

Fire Hazards Internal Fire

• As described in PSAR Section 6.1.1.8, internal fire hazards will be fully evaluated at the OL stage

• Fire protection features will be developed to satisfy DBA assumptions and PRA results and insights

• Nuclear island design consistent with RG 1.189 (PSAR Section 5.3.1.3)

• Design features include sodium leak containment, inert environments, combustible material control, fire detection, and dry-gas/water suppression systems

• Internal fire hazard reasonably characterized

• More discussion on internal fire during later presentation External Fire

• Site layout and cleared zones provide protection (PSAR Section 2.3.3.1.7)

• Wildfires are not a credible threat to site facilities

• Further evaluation of potential fires from NI and EI will be completed at the OL stage

• Conforms to RG 1.189 to ensure control room habitability and 72-hour remote shutdown capability

• External fire hazard reasonably characterized 11

Transportation, Volcanic, and Aircraft Impact Hazards Transportation Routes within 5 miles (rail, road, waterways, pipelines) assessed and discussed in PSAR Chapters 2 and 7 Evaluation consistent with RG 1.78 and RG 1.91 No chemicals or accidents identified that constitute a design-basis hazard Transportation hazard reasonably characterized Volcanic Volcanic hazards near the KU1 site assessed in PSAR Section 2.7 Site located at sufficient distance from potential eruption sources; no lava flows or pyroclastic hazards expected (Topical Report NAT-3226-A)

Tephra fall hazards evaluated and documented in staff SE (ML24198A093)

Volcanic hazard reasonably characterized Aircraft Impact PSAR Table 3.1-2 shows aircraft impact accidents quantitatively screened out of the PRA Staff evaluation of aircraft impact analysis and facility effects provided in SE Section 11.4 Aircraft impact hazard reasonably characterized 12

Staff Findings

• Potential hazards are reasonably characterized at the CP stage

• SSC designs and mitigation strategies ensure required safety functions are maintained

• Further detailed evaluations will be performed at the OL stage for high-risk hazards (internal & external fires, internal & external floods, high winds, tornadoes, missiles, seismic, volcanic, and aircraft impact)

• Internal and external hazards PRAs to be fully developed at OL stage in conformance with non-LWR PRA standard and RG 1.247 13

Seismic Classification and Seismic Interaction 14 Safety-Related SCS1 SCS2 Non-Safety-Related with Special Treatment SCN1 SCN2 SCN3 Seismic Interaction Physical Interaction Source Sodium Fire Interaction Source

• Original classification approach did not fully consider seismic common cause failure or limit state relative to SSC function

• Challenges of seismic classification prior to seismic PRA (SPRA) being available

• Revised approach to seismic classification provides reasonable assurance that performance targets will be met at the OL

Safety-Related Seismic Classification and Design

• SCS1, SR seismic risk significant, is initially assigned to all SR SSCs

• If failure of the SR SSC in a seismic event does not adversely affect the safety significant function, an SCS2 classification may be assigned

• Both SCS1 and SCS2 are designed in a manner consistent with the Seismic Category I classification in RG 1.29, using

• ASCE/SEI 4-16, Seismic Analysis of Safety-Related Nuclear Structures

• ACI 349-13, Building Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary

• ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities

• SCS2 qualification performed consistent with commercial standard 15

Non-Safety-Related with Special Treatment Classification

• SCN3 initially assigned to all NSRST SSCs

• Elevated based on

• Minimum building code requirements

• Contribution to mitigating LBEs

• Potential unmitigated dose consequences

• SPRA will be used to confirm or adjust seismic classifications at OL 16 SCN3 SCN2 SCN1 Active safety function Seismic risk-significant SSC Non-seismic risk-significant structure

Non-Safety-Related with Special Treatment Design

• NSRST seismic design parameters determined using commercial codes and standards with additional margin based on risk-significance

• IBC 2021, International Building Code

• ASCE/SEI 7-16, Minimum Design Loads and Associated Criteria for Buildings and Other Structures 17 SCN1 SCN2 SCN3 Seismic risk category IV III N/A Importance factor 1.5 1.25 1.0 Seismic qualification Yes Yes No

• ACI 318-19, Building Code Requirements for Structural Concrete

• ANSI/AISC 360-16, Specification for Structural Steel Buildings

• Margin added to fragility parameters provide reasonable assurance that SCN1 and SCN2 SSCs will remain functional at the DBHL

Seismic Interaction - Physical Interaction

• Area-based review process to identify target - source pairs

• Targets include SR and seismic risk significant NSRST SSCs

• Sources may be NSRST or NST (SR SSCs already designed to highest seismic standards)

• ASCE/SEI 7-16 used as code of record

• Design parameters consistent with SCN1

• ASCE/SEI 43-19 evaluations performed to ensure

• Inelastic deformation does not exceed limit state B

• Separation is adequate based on methods in section 7.3 of ASCE/SEI 43-19

• Barriers may also be used to protect target SSCs 18

Seismic Interaction - Sodium Fire Interaction

• Applied to all sodium containing SSCs, including some NST SSCs

• Target performance level equivalent to SR SSCs

• For NST SSCs with seismic interaction requirements,

• Programmatic controls are applied at a level consistent with NSRST SSCs

• Programs applied include the QAPD, the Post-Construction Inspection, Testing, and Analysis Program (PITAP), and the Natrium Maintenance Program 19

Seismic Isolation System

• Series of spring and damper units below reactor support structure (RSS)

• Design and qualification methodology provided in Reactor SIS Qualification Topical Report (ML25195A156), uses

• NUREG/CR-7253, as applicable

• ASME BPVC Section III, Div. 5, Subsection HF for design

• ASME QME-1 for qualification

• ASME NQA-1 for quality assurance

• ASME BPVC,Section XI, Div. 2 for monitoring, inservice inspections, and surveillance

• Limited design information available for RES umbilicals, will be reviewed at OL 20

Special Treatments for Safety-Significant SSCs

• All SR and NSRST SSCs are safety-significant and special treatments are selected to ensure they can fulfill their safety functions

• Special treatments may include:

• Selection of a nuclear code and standard

• Adding requirements beyond what is required by the commercial code (e.g., additional inspections and testing)

• Programmatic special treatments 21

• RISC definitions in 10 CFR 50.69

Design of SR Mechanical SSCs

• Codes and Standards

• ASME Section III, Division 1 for low temperature service

• ASME Section III, Division 5 for elevated temperature service

• Guard vessel designed to ASME Section VIII, Rules for Construction of Pressure Vessels with supplemental design requirements

• Materials

• Primary material of construction is austenitic stainless steel (316/316H or 304/304H)

• 304H or 316H stainless steel with ER16-8-2 weld metal for most SSCs in contact with high temperature sodium

• Other materials include 2.25Cr-1Mo, Alloy 718 and Alloy 800H

• Allowable stresses for elevated temperature limited to III-5 extended to 500,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 22

Design of NSRST Mechanical SSCs

• Codes and Standards

• ASME Section VIII for pressure vessels

• ASME B31.1 Power Piping or B31.3 Process Piping for piping systems

• NSRST design special treatments

• ASME Section III design transient loads are considered as loading input

• Apply ASME Section III, Division 5 rules or code case 2843-3 to address creep effects for components designed to Section VIII with non-negligible creep

• Apply ASME Section III, Division 5 HCB-3634 to address creep effects for piping designed to ASME B31.1 or B31.3 with non-negligible creep

• Other special treatments applied on a case-by-case basis and discussed in Chapter 7 on SSC design 23

Programmatic Special Treatments 24 Program Purpose SR Special Treatments NSRST Special Treatments Quality Assurance (QA)

Program Provides QA requirements for the design and construction of SSCs 10 CFR 50 Appendix B Graded approach Design Reliability Assurance Program (D-RAP)

Provides processes for ensuring SSC reliability during plant design phases Applied for SR and NSRST Post-Construction Inspection, Testing and Analysis Program (PITAP)

Conducts tests to demonstrate the plant constructed as designed Testing performed per 10 CFR 50 Appendix B Testing tailored to the capability assigned Mechanical Equipment Qualification Program (EQ)

Provides requirements for equipment qualification to ensure SSC performance ASME QME-1 Same as SR Comprehensive Vibration Assessment Program (CVAP)

Provides vibration assessment of reactor internals RG 1.20 Same as SR Inservice Testing Program (IST)

Provides in-service testing requirements to assess SSC performance ASME OM-2 Same as SR Reliability and Integrity Management Program (RIM)

Provides life cycle management of passive SSCs Application considers safety classification, degradations, failure consequences, reliability targets and other attributes per ASME XI-2 Only for mechanical SSCs

Reliability and Integrity Management (RIM) Program

• Scope: passive SR and NSRST mechanical SSCs

• Involves performing a degradation mechanism assessment (DMA), establishing quantitative reliability targets and implementing RIM strategies such as in-service inspection and monitoring to demonstrate reliability targets can be met

• Following process in ASME Section XI, Division 2 (XI-2) as endorsed in RG 1.246

• USO identified an R&D item associated with the development of information needed to provide assurance of adequate structural material performance for safety-significant SSCs included in the RIM program

• Discussed more in later presentation on Reactor Enclosure System (RES) 25

Staff Evaluation of RIM Process

• Preliminary RIM SSC scope is reasonable and consistent with XI-2

• Demonstrating that the reliability targets are met by RIM strategies needs to be addressed at OL

• Preliminary RIM strategy development approach appears to indicate some SSCs with lower risk of degradation would not require any monitoring or inspection

• NRC staff expects any SSC in RIM with potentially active degradation should be covered by either a primary or expansion RIM strategy demonstrated to achieve and maintain the specified reliability target for that SSC

• Based on following an endorsed process, review of the preliminary RIM documentation and the R&D item, the staff finds the applicants preliminary RIM program acceptable to support the design and construction stage 26

Acronyms 27 10 CFR - Title 10 of the Code of Federal Regulations ACRS - Advisory Committee on Reactor Safeguards ACI - American Concrete Institute AISC - American Institute of Steel Construction ANSI - American National Standards Institute ASCE - American Society of Civil Engineers ASME - American Society of Mechanical Engineers BPVC - Boiler and Pressure Vessel Code CP - Construction Permit CVAP - Comprehensive Vibration Assessment Program DANU - Division of Advanced Reactors and Non-Power Production and Utilization Facilities DBA - Design Basis Accident DBHL - Design Basis Hazard Level Div. - Division DMA - Degradation Mechanism Assessment D-RAP - Design Reliability Assurance Program EI - Energy Island EQ - Equipment Qualification Program HF - Human Factors IBC - International Building Code (IBC)

ISG - Interim Staff Guidance IST - Inservice Testing Program KU1 - Kemmerer Unit 1 LBE - Licensing Basis Event LMP - Licensing Modernization Project LWR - Light Water Reactor MIRSS - Modular Isolated Reactor Support Structure MPH - Miles Per Hour NEI - Nuclear Energy Institute NI - Nuclear Island NQA - Nuclear Quality Assurance Standard NRC - Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation NSRST - Non-safety-related with Special Treatment NST - No Special Treatment N/A - Not Applicable OL - Operating License PDC - Principal Design Criteria PITAP - Post-Construction Inspection, Testing, and Analysis Program PRA - Probabilistic Risk Assessment PSAR - Preliminary Safety Analysis Report QA - Quality Assurance QAPD - Quality Assurance Program Description QME - Qualification of Active Mechanical Equipment RES - Reactor Enclosure System RG - Regulatory Guide RISC - Risk-Informed Safety Class RIM - Reliability and Integrity Management Program RXB - Reactor Building R&D - Research and Development SE - Safety Evaluation SEI - Structural Engineering Institute SPRA - Seismic Probabilistic Risk Assessment SR - Safety Related SSC - Structures, Systems, and Components UAL1 - Advanced Reactor Licensing Branch 1 UTB2 - Advanced Reactor Technical Branch 2 USO - US SFR Owner