ML25293A453

7 - NRC Review of Cross-Cutting Systems

ML25293A453
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	10/20/2025
From:	Reed Anzalone, Matthew Hiser, Nick Melly, Jay Robinson NRC/NRR/DRA, Advisory Committee on Reactor Safeguards, NRC/NRR/DANU, NRC/RES/DRA
To:
References
Download: ML25293A453 (1)

Text

NRC Review of Kemmerer Unit 1 Construction Permit Application Sodium Fire and Leak Mitigation Nicholas Melly (RES/DRA), Matt Hiser (NRR/DANU),

Jay Robinson(NRR/DRA), Reed Anzalone (NRR/DANU)

ACRS Subcommittee Meeting October 22-23, 2025

Topics

• Conventional Fire Protection System Design (SE section 7.5)

• Quality Standards / Natural Phenomena / Environmental and Dynamic Effects

• No water-based suppression in the RAB and RXB

• Sodium Fire Protection System Design(SE section 8.3)

• Cells / Guard Vessels / Pipes / Enclosures

• Catch Pans /Leakage Detection / Materials

• Fire Areas / Barriers

• Sodium Leakage Concepts & Scenario Postulation

• Fire Analysis 2

Overview of Fire Protection System Review

• The PSAR commits to meeting 10 CFR 50.48(a), PDC 3, and Regulatory Guide (RG) 1.189, Revision 5, which will ensure the design maintains adequate defense-in-depth protection against conventional fire hazards

• This should address prevention, detection, containment, suppression, and the ability to maintain safe shutdown

• The KU1 design uses sodium fast reactor industry information to develop Natrium-specific guidance, requirements, and criteria 3

Conventional Fire Protection System Design Quality Standards / Natural Phenomena / Environmental and Dynamic Effects 4

• PSAR & Supporting Calculations

• PSAR table 1.4 Full conformance with RG 1.189

• PSAR table 1.4 Fire Protection/Building Construction Codes and Standards

• PSAR section 6.1.3.1 - Sodium-containing SSCs are treated as a special seismic interaction source based on the potential for seismic-induced fire and are required to demonstrate the ability to retain sodium inventory following an SSE

• PSAR section 7.5.2.2 - NFP components are designed to withstand the effects of natural phenomena without loss of capability to perform their safety functions. PSAR section 3.1 describes PRA and risk insights

• PSAR section 8.2.1 - There are no water-containing systems within the RXB and RAB to avoid contact between sodium and water and to limit the adverse effects of chemical reactions between sodium and water or the capability of SSCs to perform safety-significant functions

• PSAR table 3.1 Fire PRA will be available with the OL application (discussed in ACRS Oct 8th 2025)

Sodium Fire Protection System Design Approach

• The fire protection program will be fully developed at the operating license stage to meet 10 CFR 50.48(a) using the guidance in RG 1.189, Revision 5

• Large sodium fires are passively mitigated through NNA design features

• No active or manual fire suppression for large sodium fires and sodium leaks

• Deep-seated sodium fires contained within NNA features and passively self-suppressing

• Firefighting strategies for conventional deep-seated fires will be described in the fire protection plan submitted at the OL stage

• Fire water suppression systems are not provided in areas with sodium-containing systems in the FHB to meet PDC 74 related to sodium-water reactions

• Due to lack of water-based suppression, conventional deep-seated firefighting strategies may require offsite response with specialized training 5

Sodium Fire Mitigation Design Features

• Various methods of limiting sodium interactions with air and concrete are identified, including:

• Seals and inert gases,

• Cells,

• Guard pipes,

• Guard enclosures and clamshells,

• Catch pans,

• Inerted guard vessels,

• Leakage detection,

• Less combustible materials, and

• Fire areas and barriers 6

Design Features: Cells and Guard Vessels 7

• Cells: rooms with a limited leak rate and known oxygen volume

• Either actively inerted or become de facto inerted upon consumption of the limited available oxygen after a sodium leak

• For cells without an inerted environment, the cell will either be evaluated for combustion or a suppression deck will be included

• Guard vessels

• Used to prevent sodium reactions in the event a leak from another vessel containing sodium

• Annulus space is inerted and contains leak detection

• Staff evaluation

• The staff notes that cells and guard vessels are generally an effective means of preventing or mitigating sodium-air reactions and are reasonable to use in various areas of the plant

Design Features: Guard Pipes and Enclosures 8

• Guard Pipes

• May be inerted or not but act as a full secondary pressure boundary to contain activated sodium

• Guard enclosures:

• Similar to guard pipes, with the addition of a drainage to an acceptable location (e.g. drain tank cell or catch pan)

• Staff evaluation

• The staff notes that guard pipes and enclosures are generally an effective means of preventing or mitigating sodium-air reactions and are reasonable to use in various areas of the plant based on the risk of sodium leakage and fire Guard enclosure example (Illustrative only-not actual design drawings)

Design Features: Clamshell Enclosures 9

• Clamshell as a form of guard enclosure used where inspections are expected

• IHT SSCs located within the HAA and form the fire area and boundary between the RAB pipe chase room and the HAA

• Ensure pressurized sodium in the clamshell cannot spray into the HAA

• This design is noted as a "first-of-a-kind" and aspects of the clamshell design will be tested to ensure it can perform its required function

• Testing methods and acceptance criteria to be identified in the OL application to ensure effectiveness when exposed to design temperatures and pressures

• Staff evaluation

• Proposed use of clamshell enclosures to prevent sodium leakage from IHT piping from entering the HAA and causing a sodium fire in a critical area of the plant is very important

• USO plans to perform testing of the clamshell design as well as leak-tightness and pressure testing of these SSCs after installation

• NRC staff will review the final design at to OL stage, including configurationally-dependent PRA inputs for fire scenarios

Design Features: Catch Pans 10

• Steel-lined volumes designed to retain the design-basis sodium leak volume plus margin and are used as drain locations that collect leaked sodium from guard enclosures or sloped floors

• Staff observes the use of catch pans is generally an effective means of mitigating sodium-air reactions and is consistent with consistent with PDC 3 and 73

• Final catch pan design and locations will be evaluated during the OL

• Design incorporates margin as described in ANS 54.8 with additional safety factor (Illustrative only-not actual design drawings)

Leakage Detection 11

• Leakage detection equipment for liquid sodium, sodium aerosols, and reaction products, is provided to inform plant operators in the event of a leak

• Provided at multiple points between pipe supports along all normally sodium containing systems (i.e., drain lines excepted), within either the guard pipe, guard enclosure, or pipe insulation

• Leak detectors are capable of detecting leaks within the entire length of pipe it covers

• Conventional fire detection is installed in non-inerted areas to provide supplemental sodium leak detection

• The fire protection program supports three NSRST functions that isolate SSCs in the event of a sodium leak. These are:

• DL4-RR3a, SPS Supply Valve Isolation on Leak Detection (NSRST)

• DL4-RR3b, SPS Pump Trip on Leak Detection (NSRST)

• DL4-RR3c, SPS Cell Barrier Isolation on Leak Detection (NSRST)

Sodium Leakage Concepts

• PSAR states that leaks are postulated for all piping that normally contains sodium

• Design basis (DB) leakage

• If within a process cell or guard enclosure, leakage will drain to a catch pan

• Non-inerted: analyzed for sodium-air reactions to inform design features

• Inerted: Some process cells have an inerted environment and will not include a suppression deck

• DB leakage outside of guard enclosures or process cells not addressed in PSAR section 8.2.2

• Beyond design basis (BDB) leakage

• Evaluated as part of the integrated risk assessment considering worst case, maximum leak sizes

• Additional leak mitigation features may be added to address unacceptable consequences in the OLA

• Staff evaluation

• Conceptual approach reasonable but preliminary with limited detail

• Limited description of how DB and BDB leakage will be established and analyzed 12

Sodium Leakage Postulation

• Sodium leak postulation approach adapted from BTPs 3-3 and 3-4* for LWRs

• For high stress areas, DB leakage crack sizes assumed to be Dt/4 (diameter multiplied by thickness divided by 4) crack for seismically qualified piping

• For lower stress areas, postulated crack sizes will use a methodology to be provided in the OLA

• Staff evaluation

• Sodium leak postulation from vessels will need to be justified at the OL

• Staff expects DB leaks to be postulated from any sodium-containing SSC with a single barrier, including vessels

• BTP 3-3 and 3-4 are specific to LWRs

• Even small sodium leaks can create a fire hazard if not mitigated

• Significant differences between sodium systems and LWRs (e.g., pressure, temperature, coolant)

• Conceptually, graded approaches are a reasonable way to analyze and address sodium leakage

• Specific criteria and basis not sufficiently developed and justified at the CP stage 13

• Branch Technical Positions (BTPs) 3-3 and 3-4 from NUREG-0800 for moderate energy piping for light water reactors (LWRs)

Fire Areas / Barriers/ Fire Analysis 14

• Fire areas are established so redundant trains of SSCs are in separate fire areas per the guidance in RG 1.189

• Separation by physical barriers / selective positioning

• Fire scenarios will be modeled and analyzed with sensitivity analyses

• Final design of fire areas and barriers will be reviewed at the final design stage in the OL review

• Fire modeling, fire PRA, fire hazards analysis, and fire safe shutdown analysis will be performed as part of the OL application

• Strategies may include manual firefighting, automatic or manual suppression with agents compatible with sodium, inerting, purging, using incombustible materials, and SSC specific design features

• Sodium fire modeling methodology with complete verification and validation and inputs to support the fire PRA will be developed at the OL stage to meet ASME/ANS RA-S-1.4-2021

Conclusion KU1 hazards assessment is reasonable for CP application and consistent with PDC 3, RG 1.189 and applicable codes and standards Staff expectations for USO:

• Validate all assumptions during OL stage

• Continue to develop and prescribe suppression techniques to be used in the event of a fire in sodium containing areas for both conventional and sodium involved fires

• Perform methodology development including Verification and Validation approach for sodium fires

• Perform fire PRA to inform design characteristics and appliable hazards

• Verify the clamshell design will be tested to ensure it can perform its required function with testing methods and acceptance criteria 15

Acronyms 16 ACRS - Advisory Committee on Reactor Safeguards ANS - American Nuclear Society ASME - American Society of Mechanical Engineers BDB - Beyond Design Basis BTP - Branch Technical Position CFR - Code of Federal Regulations CP - Construction Permit DANU - Division of Advanced Reactors and Non-power Production and Utilization Facilities DB - Design Basis DID - Defense In Depth DRA - Division of Risk Analysis FHB - Fuel Handling Building HAA - Head Access Area IHT - Intermediate Heat Transport System RAB - Reactor Auxiliary Building RES - Reactor Enclosure System RG - Regulatory Guide RXB - Reactor Building SE - Safety Evaluation SFR - Sodium Fast Reactor SPS - Sodium Processing System SR - Safety Related SSC - Structures, Systems, and Components SSE - Safe Shutdown Earthquake USO - US SFR Owner KU1 - Kemmerer Unit 1 LWR - Light Water Reactor NEI - Nuclear Energy Institute NNA - Sodium Leak Detection, Collection, and Containment System NFD - NI Fire Water Distribution System NFP - NI Fire Protection System NI - Nuclear Island NRC - Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation NSRST - Non-safety-related with Special Treatment NST - No Special Treatment OL - Operating License PDC - Principal Design Criteria PRA - Probabilistic Risk Assessment PSAR - Preliminary Safety Analysis Report

NRC Review of Kemmerer Unit 1 Construction Permit Application Instrumentation and Control Systems Calvin Cheung & Dinesh Taneja NRR/DEX/EICB & ELTB ACRS Subcommittee Meeting October 21-23, 2025

Staff Review of I&C Systems Design

• Primarily followed DRG guidance for evaluation of I&C system design for the Kemmerer Unit 1 CP application

• Evaluated I&C system architecture for adherence to the fundamental I&C design principles (i.e., independence, redundancy, diversity in support of DID, and deterministic behavior (repeatability and predictability))

• Focused review on SR and NSRST functions performed by the I&C systems 18

Staff Summary of Review for I&C System

• Preliminary design details are commensurate to the level of details necessary for CP.

• Commitments to meeting the regulatory requirements are found to be adequate.

• Appropriate Consensus standards are cited in the PSAR for the I&C systems design

• IEEE Std 603-2018 is being used for the KU1 I&C system design. However, the application commits to demonstrating compliance to 10 CFR 50.55a(h).

• Applicant commits to addressing the limitations and conditions of the incorporated by topical report, NAT-4950-A, I&C Architecture & Design Basis TR, Rev. 2 (ML25232A240) in the OL application as the I&C design matures.

• Information is consistent with the specified PDCs and further information as may be required to complete the review can be reasonably left for later consideration in the OL.

19

Focus Areas for Operating License

• I&C architectures adherence to fundamental I&C design principles

• Concepts of simplicity applied in I&C system design

• Allocation of plant functions to applicable I&C system defense layers

• Quantitative reliability of the I&C systems designated for each defense layer, supporting qualitative reliability arguments

• Hazard analysis

• Single failure analysis

• Digital CCF coping analysis

• Digital I&C system timing and real-time performance analysis

• Setpoint methodology

• Quality and equipment qualification

• Compliance to applicable regulations

• Security by design 20

I&C Systems Design Review Interfaces Interface final I&C design review with other technical areas, including:

• PRA

• Reactor Systems

• Human Factors

• Technical Specifications

• EQ testing with Electrical & Civil/Structural engineering

• Post-accident monitoring with TS, HFE, Rad Protection, and operator licensing 21

NRC Review of Kemmerer Unit 1 Construction Permit Application Buildings and Structures Tracy Radel & Reed Anzalone NRR/DANU/UTB2 ACRS Subcommittee Meeting October 22-23, 2025

Topics

• Reactor Building

• Fuel Handling Building

• Reactor Auxiliary Building

• Nuclear Island Control Building 23

Overview of Building Classifications 24 Nuclea r Isla nd Control Building NST NSRST SR

Reactor Building (RXB) - Overview

• Substructure (SR, SCS1): Embedded reinforced concrete structure

• Superstructure (NST, with seismic interaction requirements): above-grade steel-framed, supported on independent concrete foundation

• Substructure and superstructure are isolated from each other

• The RSS transfers load from the RES to the RXB through seismic isolators 25

RXB Substructure

• Design consistent with SCS1 seismic classification

• Use of ASCE/SEI 4-16 and ACI 349-13

• Designed to withstand DBHLs with appropriate load combination, soil structure interaction, and material considerations

• Designed to allow for inspection and testing of the RAC, HAA, and portions of the RXB

• Supports

• SR passive RAC operation by maintaining the outer boundary of the RAC air flow path (DL3-HR4)

• NSRST functional containment barrier functions through the GV, HAA walls, and HAA HVAC isolation (DL4-RR1)

• Load path for the RAC stack and the RSS

• Preliminary design consistent with PDC 1-4, 16, 35-37, 81, and 82 26

RXB Superstructure

• Seismic interaction requirements applied to protect RAC and other SR and seismic risk significant NSRST SSCs

• Design consistent with these requirements

• Use of ASCE/SEI 7-16, Risk Category IV

• ASCE/SEI 43-19 evaluations to ensure Limit State B is met and separation distances are adequate

• Supports RXB bridge crane

• If crane classification changes due to designation as a preventative measure, RXB superstructure classification may change but current design is consistent with SCN1 27

Fuel Handling Building - Overview

• Substructure (SR, SCS1): Embedded reinforced concrete structure

• Superstructure (NSRST, SCN1): above-grade steel framed with portions comprised of concrete shear walls, main building supported by FHB substructure

• Separate reinforced concrete slabs support the FHB north wing and transfer corridor

• FHB north wing houses liquid and solid radwaste processing systems within shielded reinforced concrete enclosures and the NI water system and NHV on second level

• Enclosed transfer corridor provides a path for rail-mounted fuel handling equipment to transition from RXB to FHB 28

Fuel Handling Building - Substructure

• Houses the EVST, PIC, and spent fuel pool (SFP)

• Design consistent with SCS1 seismic classification

• Use of ASCE/SEI 4-16 and ACI 349-13

• Acceptance criteria for SFP liner in accordance with ANSI/AISC N690-18

• Designed to withstand DBHLs with appropriate load combination, soil structure interaction, and material considerations

• Supports

• External hazard protection for SR and NSRST SSCs

• Load path for SR and NSRST SSCs

• Preliminary design consistent with PDC 1-4 29

Fuel Handling Building - Superstructure

• Design consistent with SCN1 requirements

• Use of ASCE/SEI 7-16, ACI 318-19, ANSI/AISC 360-16

• Supports

• NSRST function DL4-RR7 which requires the FHB superstructure to maintain a slightly negative pressure for radionuclide retention

• External hazard protection for NSRST SSCs

• Temporary load path for SR SSCs

• Load path for NSRST SSCs

• Preliminary design consistent with PDC 1-4, and 16 30

Reactor Auxiliary Building (RAB) - Overview 31 Reactor Building Reactor Auxiliary Building

• Substructure (NSRST): Embedded reinforced concrete structure with two levels

• Superstructure (NSRST): above-grade steel framed with metal siding and roof, supported on substructure concrete slab

• Connected to the RXB

• Below grade for piping access

• Above grade for personnel passage

Reactor Auxiliary Building - Design and Functions

• Design consistent with SCN1 seismic classification

• Use of ASCE/SEI 7-16, ACI 318-19, ANSI/AISC 360-16

• Designed to withstand DBHLs with appropriate load combination, soil structure interaction, and material considerations

• Supports

• SPS Cells Barrier as enveloping boundary for functional containment for radionuclide retention (DL4-RR3)

• External Hazard Protection for NSRST SSCs

• Load Path for NSRST SSCs

• Preliminary design consistent with PDC 1-4, and 16 32

Nuclear Island Control Building (NCB) -

Overview 33

• Substructure (SR): Embedded reinforced concrete structure

• Houses SR electrical systems and components, NSRST electrical equipment, and remote shutdown complex (RSC)

• Superstructure (NSRST): above-grade one-story steel framed metal building

• Houses the main control room (MCR)

Nuclear Island Control Building (NCB) -

Substructure

• Design consistent with SCS1 seismic classification

• Use of ASCE/SEI 4-16, ACI 349-13, and ANSI/AISC N690-18

• Designed to withstand DBHLs with appropriate load combination, soil structure interaction, and material considerations

• Designed to provide adequate radiation protection to permit access and occupancy of the RSC under accident conditions

• Supports

• External Hazard Protection for SR and NSRST SSCs

• Load Path for SR and NSRST SSCs

• Preliminary design consistent with PDC 1-4, and 19 34

Nuclear Island Control Building (NCB) -

Superstructure

• Design consistent with SCN1 seismic classification

• Use of ASCE/SEI 7-16, ACI 318-19, ANSI/AISC 360-16

• Designed to withstand DBHLs and ensure operators remain safe and can perform safety-significant functions (beyond the level required by the IBC)

• Designed to provide adequate radiation protection to permit access and occupancy of the MCR under accident conditions

• Supports

• External Hazard Protection for NSRST SSCs

• Load Path for NSRST SSCs

• Preliminary design consistent with PDC 1-4, and 19 35

Acronyms 36 ACRS - Advisory Committee on Reactor Safeguards ACI - American Concrete Institute AISC - American Institute of Steel Construction ANS - American Nuclear Society ASME - American Society of Mechanical Engineers ASCE - American Society of Civil Engineers CCF - Common Cause Failure CFR - Code of Federal Regulations CP - Construction Permit DEX - Division of Engineering and External Hazards DID - Defense In Depth DBHL - Design Basis Hazard Level NEI - Nuclear Energy Institute NI - Nuclear Island NHV - NI Heating, Ventilation, and Air Conditioning System NRC - Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation NSRST - Non-safety-related with Special Treatment NST - No Special Treatment OL - Operating License PIC - Pool Immersion Cell PRA - Probabilistic Risk Assessment PSAR - Preliminary Safety Analysis Report RAB - Reactor Auxiliary Building DRG - Draft Regulatory Guide EICB - I&C Branch ELTB - Long Term Operations and Modernization Branch EQ - Equipment Qualification EVST - Ex-Vessel Storage Tank FHB - Fuel Handling Building GV - Guard Vessel HAA - Head Access Area HFE - Human Factors Engineering HVAC - Heating, Ventilation, and Air Conditioning I&C - Instrumentation and Controls IEEE - Institute of Electrical and Electronics Engineers KU1 - Kemmerer Unit 1 MCR - Main Control Room NCB - NI Control Building

Acronyms 37 RAC - Reactor Air Cooling System RES - Reactor Enclosure System RG - Regulatory Guide RSC - Remote Shutdown Complex RSS - Reactor Support Structure RXB - Reactor Building SE - Safety Evaluation SEI - Structural Engineering Institute SFP - Spent Fuel Pool SFR - Sodium Fast Reactor SPS - Sodium Processing System SR - Safety Related SSC - Structures, Systems, and Components TR - Topical Report TS - Technical Specifications USO - US SFR Owner