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Kemmerer Unit 1 ACRS Subcommittee Presentation October 8th, 2025
	ML25280A230
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Site:	Kemmerer File:TerraPower icon.png
Issue date:	10/08/2025
From:	Williams E; TerraPower
To:	; Office of Nuclear Reactor Regulation
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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved a TerraPower & GE Vernova Hitachi Nuclear Energy technology Natrium Design Overview for Kemmerer Unit 1 ACRS Meeting Eric Williams TerraPower EVP & Chief Operating Officer October 8, 2025

Sodium-cooled fast reactor technology using inherent and passive safety features

Reduced quantity of nuclear-grade safety equipment and materials

Molten salt integrated energy storage system for grid flexibility

Address energy demand with a safe, reliable, and efficient design

Simplify and streamline the plant design

Integrate with other diverse sources of generation on the grid Objectives Key Features Introduction to the Natrium Reactor 2

TerraPowers Mission: to solve the worlds toughest problems in energy, climate, and human health through innovative nuclear technology.

Licensing basis methodology uses an integrated decision-making process based on NEI 18-04 as endorsed by Reg Guide 1.233 Selection of licensing basis events (LBE)

Safety classification of structures, systems, or components (SSC)

Application of special treatments Determination of defense-in-depth (DID) adequacy

3 Fundamental Safety Functions (FSFs) ensure cumulative risk objectives of regulatory dose criteria and Quantitative Health Objectives are met

Defense Line (DL) strategy provides independent and redundant mitigation strategies that ensure FSFs are satisfied

Specified acceptable system radionuclide release design limits (SARRDL) are established to:

Ensure most limiting LBE does not exceed siting regulatory dose limits Ensure 10 CFR 20 dose limits to the public are not exceeded Plant Safety Overview

Retaining Radionuclides: performed using a functional containment strategy

Uses diverse passive barriers

Controlling Heat Generation: performed by inserting control rods into the reactor core and has active, passive, and inherent means of being achieved across DLs

2 diverse means of inserting 2 control rod banks of diverse geometrical design

Controlling Heat Removal: has active, passive, and inherent means of being achieved across the DLs

2 diverse residual heat removal systems, Intermediate Air Cooling System (IAC) and Reactor Air Cooling (RAC)

Fundamental Safety Functions

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Control Motor-driven control rod runback and scram follow Gravity-driven control rod scram Inherently stable with increased power or temperature Cool In-vessel primary sodium heat transport (limited penetrations)

Intermediate air cooling natural draft flow Reactor air cooling natural draft flow - always on Contain Low primary and secondary pressure Sodium affinity for radionuclides Multiple radionuclides retention boundaries Safety Features

Simplified Response to Abnormal Events

Reliable reactor shutdown

Transition to coolant natural circulation

Indefinite passive emergency decay heat removal

Low pressure functional containment

No reliance on Energy Island for safety functions

No Safety-Related Operator Actions or AC power

Pool-type Metal Fuel SFR with Molten Salt Energy Island

Metallic fuel and sodium have high compatibility

No sodium-water reaction in steam generator

Large thermal inertia enables simplified response to abnormal events

Technology Based on U.S. SFR Experience

EBR-I, EBR-II, FFTF, TREAT

SFR inherent safety characteristics demonstrated through testing in EBR-II and FFTF Safety Features 6

Reactor Aux. Building Intermediate Sodium Hot Leg Intermediate Sodium Cold Leg Reactor and Core Intermediate Air Cooling Head Access Area Refueling Access Area Reactor Air Cooling / Reactor Cavity Reactor Building Fuel Handling Building Reactor Air Cooling Ducts Spent Fuel Pool (water)

Sodium Int. Loop Sodium/Salt HXs Salt Piping to/from Thermal Storage System Ground Level Natrium Cross-Section

Thermal energy storage tanks

Steam generators

Feedwater

Condenser

Turbine

Balance-of-plant supporting systems

Reactor

Sodium systems

Reactor supporting systems

Spent fuel systems

Radwaste systems

Structures Nuclear Island (NI)

Energy Island (EI)

Nuclear Island & Energy Island Systems 8

Reactor Core System (RCC): includes fuel, control, reflector, shield; designed to sustain a nuclear chain reaction for fission heat generation

Reactor Air Cooling System (RAC): provide passive decay heat removal to the environment through natural convection of air

Primary Heat Transport System (PHT): includes hot pool, cold pool, warm pool, heat exchangers, and pumps; functions is to transport heat from the reactor core to intermediate sodium

Intermediate Heat Transport System (IHT): includes piping, pumps, and sodium-salt heat exchangers; functions to transport heat away from PHT and to the salt system

Intermediate Air Cooling System (IAC): includes blowers and sodium-air heat exchangers; transfers heat from the IHT to the atmosphere in active or passive mode Nuclear Island Systems

Sodium Cover Gas System (SCG): consists of gas distribution, monitoring, and filtering; functions to control, monitor, and supply inert argon gas to various systems and components throughout the Nuclear Island

Sodium Processing System (SPS): monitors and removes contaminants (cesium) and compounds (oxygen, hydrogen) from primary, intermediate, and Ex-Vessel Storage Tank liquid sodium inventory

Control Rod Drive System (CRD): provides core reactivity control for core power control, power shaping, and reactor shutdown (motor and gravity driven)

Instrumentation & Control (I&C): provides manual and automatic control of systems during normal and transient conditions, includes reactor protection, nuclear instrumentation, radiation monitoring, reactor instrumentation, and others Nuclear Island Systems

In-Vessel Fuel Handling System: manipulates the positions of core assemblies during refueling operations

Ex-Vessel Fuel Handling System: facilitates the receipt, conditioning, storage, installation, and removal of all core assemblies

Water Pool Fuel Handling System: stores irradiated core assemblies within the water-cooled pool, ensures decay heat is rejected, and maintains water chemistry

Gaseous/Liquid/Solid Radioactive Waste Systems: process gaseous, liquid, and solid radioactive wastes for release and disposal

Sodium Leak Detection, Collection, and Containment System (NNA): detects, collects, and contains sodium leaks to mitigate sodium fires using guard enclosures/clamshells, catch pans, and leak detectors Nuclear Island Systems

Reactor Building (RXB): houses reactor and associated SSCs that support reactor operation

Fuel Handling Building (FHB): houses plant operations and equipment related to fuel receipt, refueling operations, storage, spent fuel pool, and radwaste systems

Reactor Auxiliary Building (RAB): houses critical functions and processes that support IHT, sodium-salt heat exchangers, intermediate sodium pumps, and SPS

NI Control Building (NCB): houses the Main Control Room and associated equipment and reactor protection system vaults and other safety-significant systems and equipment

Fuel Auxiliary Building (FAB): provides centralized location for employee and visitor ingress to, and egress from major NI facilities Nuclear Island Structures

Nuclear Island, Thermal, and Energy Island Salt Systems (NSS, TSS, ESS):

removes heat from intermediate sodium for thermal storage and energy production in Power Cycle Systems

Power Cycle Systems: removes heat from the TSS to generate steam in the steam generators using high pressure feedwater, rotate the turbine and generator, then condense in the condenser using cooling towers

Main Power: 3 main functions

Transfer power between Kemmerer Unit 1 and 230 kV transmission lines

Transfer power from generator to 230 kV switchyard and supply plant auxiliary loads via unit auxiliary transformer during normal operation

Transfer power from the 230 kV switchyard to plant auxiliary loads using reserve auxiliary transformer when normal main power unavailable Energy Island Systems

1. Primary sodium

2. Intermediate sodium

3. Nitrate salt

4. Water for steam generation Natrium Operational Interface 15 Nuclear Island Energy Island 1

2 3

3 4

Independence of operation of the NI and EI is a key aspect of the Natrium design philosophy

Steam generation is independent of reactor power

Thermal inertia of IHT and PHT are such that any changes in EI salt conditions can be adequately responded to using only signals monitored within the NI

Heat removal is ensured by the RAC and IAC systems of the NI

Critical for application of design codes to the Energy Island and applicability of NRC regulations

Influences the construction and licensing approaches Nuclear & Energy Island Design Interface

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved a TerraPower & GE Vernova Hitachi Nuclear Energy technology Questions?

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved a TerraPower & GE Vernova Hitachi Nuclear Energy technology Kemmerer Unit 1 ACRS Subcommittee Meeting Licensing Modernization Project Topics October 8, 2025

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved a TerraPower & GE Vernova Hitachi Nuclear Energy technology LBE Selection

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved The scope of hazards considered for screening was identified through review of RG 1.247 Tables B-1 and B-2 and the PRISM PRA Hazard Identification Analysis.

Qualitative and Quantitative screenings are based on generic screening criteria specified in ASME/ANS RA-S-1.4-2021.

Qualitative Screening

Screening Criterion SCR-3

Supporting Requirement HS-B5 Quantitative Screening

Screening Criterion SCR-1 or SCR-2

Supporting Requirements HS-C6 and HS-C7 Hazards that do not screen qualitatively or quantitively proceed to detailed PRA analysis.

Hazard Screening 3

In accordance with Appendix A of RG 1.253, this is an appropriate scope of the PRA at the CP stage.

PRA for internal and external hazards that were not screened out will be developed at the OL stage.

PRA Scope 4

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Performed in accordance with NEI 18-04, as endorsed by RG 1.233 IE selection Review of published IE lists for LWRs from the EPRI and NRC, including:

NUREG/CR-5750, NUREG/CR-6928, and NUREG/CR-3862

EPRI report 3002003129 Review of published IE lists for non-LWRs, including:

Experimental Breeder Reactor II (EBR-II)

KALIMER-600

Prototype Fast Breeder Reactor (PFBR)

Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID)

Japanese Liquid Metal Fast Breeder Reactor (LMFBR)

PRISM

Advanced Liquid Metal Reactor (ALMR)

Review of Natrium Plant Event Sequences list System and FMEA reviews LBE Selection 5

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved IE grouping and IE frequency assignment Generic data sources, design specific information, and fault tree modeling Event Sequence analysis Determines the combinations of POSs, IEs, safety functions, system failures and successes, and end states that may involve a release of radioactive material Develops event trees for scenarios that can occur following the occurrence of each initiating event Event Sequence grouping to Event Sequence Families Grouped based on shared POSs, IEs, challenges to plant safety functions, and end states Event Sequence Families are further grouped into LBEs Grouped to reduce redundant LBEs LBE Selection 6

DBAs assume that only SR SSCs and passive SSCs not impacted by the initiating event are available to mitigate postulated event sequence consequences to within the 10 CFR 50.34 dose limits.

At CP stage, DBAs are derived from:

DBEs At OL stage, DBAs will be derived from:

DBEs

AOOs with 5th percentile frequency less than 1x10-2/plant-year

BDBEs with 95th percentile frequency greater than 1x10-4/plant-year LBE Selection 7

Reference:

NEI 18-04 Rev. 1

/plant-year OQEs considered for cliff edge effects include:

Events that have frequencies, including 95th percentiles, between 5x10-7/plant-year and 1x10-7/plant-year Events that could have a high consequence (30-day TEDE dose above 1000 rem at the EAB)

Evaluation of frequency and dose margins for OQEs with 95th percentile frequencies in the BDBE region Evaluation of the highest consequence event quantified in the PRA model Evaluation of Cliff Edge Risk 10

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved a TerraPower & GE Vernova Hitachi Nuclear Energy technology Safety Functions and Classification

Reference:

NEI 18-04 Rev. 1

Reference:

NEI 18-04 Rev. 1

Reference:

NEI 18-04 Rev. 1

PRA Safety Function identification numbers indicate which fundamental safety function and layer of defense the function is associated with.

Example: DL3-HR4 (Inherent - RAC Operation) is a defense layer 3 (safety-related) heat removal function.

Application of an inspection program is an example of a DL1 feature.

DL2 is the normal response to most postulated initiating events. DL2 is typically NST.

Heat removal via Intermediate Air Cooling (IAC) system in active mode (classified as NST).

DL3 is typically sufficient for all DBAs/DBEs. DL3 is SR.

Heat removal via Reactor Air Cooling system (RAC) (classified as SR).

DL4 is for BDBEs or because further defense is required. DL4 is typically NSRST.

Heat removal via IAC in passive mode (classified as NSRST for DID).

DL5 is related to emergency planning.

PSF Classification Example - Cool Reactor Core 17

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Supporting functions Receive the safety classification of the functions they support where the support function failure would cause failure of the supported function Temporary load path support functions may be assigned a lower safety classification than the functions they support because they are in service for a limited amount of time (e.g., travel paths for SR fuel handling vessels).

Preventative measures Are capable of preventing postulated initiating events from progressing to transients modeled in the PRA A set of SSCs are classified as NSRST for DID adequacy as performing a safety-significant preventative measure for each PIE that exceeds a defined dose criterion.

Safety Classification 18

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