Text

Uso ACRS Subcommittee Meeting Slides Oct 8-Oct 9 (002)
	ML25282A002
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	10/08/2025
From:	; TerraPower
To:	; Office of Nuclear Reactor Regulation
References
	DE-NE0009054
	Download: ML25282A002 (1)
	v • d • e

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

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 9, 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 PRA Methodology

Developed to meet the applicable requirements of ASME/ANS RA-S-1.4-2021 as endorsed by RG 1.247 to implement NEI 18-04 as endorsed by RG 1.233.

Self-assessment concluded that the PRA is acceptable to use for the CPA.

Conducted using the guidance from NEI 20-09 as endorsed by RG 1.247

Utilized Table A-2 and Table A-3 of RG 1.253 Revision 0 to establish acceptability of the PRA at the CP stage

An evaluation comparing the CP stage PRA to the CP stage design confirms the PRA reflects the design.

PRA Methodology 22

Overviews of the PRA elements that were reviewed by the self-assessment were added to Section 3.1.1 of PSAR Revision 1 PRA Methodology 23 PRA Element PRA Element Plant Operating State Analysis Data Analysis Initiating Events Analysis Hazard Screening Event Sequence Analysis Event Sequence Quantification Analysis Success Criteria Analysis Mechanistic Source Term Analysis Systems Analysis Radiological Consequence Analysis Human Reliability Analysis Risk Integration

PRA Configuration Control process ensures the PRA is maintained and updated to reflect the as-designed, as-to-be-constructed, and as-intended-to-operate plant design.

Model Change Tracking Process

Monitors changes to design and operation of the plant, PRA technology, PRA data inputs, industry experience, and plant performance

Evaluates the significance of potential design changes on the PRA

Tracks the incorporation of the change into the PRA PRA Methodology 24

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 Functional Containment

Barrier or set of barriers that effectively limits transport of radioactive material to the environment.

Barrier type defined by function

Primary barrier: SSC that performs radionuclide retention function necessary to keep offsite DBA doses within regulatory limits or keep DBE doses from exceeding F-C targets.

Enveloping barrier: SSC that provides a backup radionuclide retention function following failure or breach of an associated primary barrier.

Definitions 26 Functional containment barriers are identified as physical system boundaries or structures for which leakage performance can be specified in design and verified by testing or associated analysis.

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Maintain at least one barrier, beyond fuel cladding, of suitable performance, reliability, and classification between a source and the environment.

Determine the number of barriers between a source and the environment having specific performance criteria and requiring additional maintenance and testing.

Provide a framework for iterations between source term analysis, design, and PRA to optimize the selection and classification of barriers.

Overall Approach and Strategy 27

Reference:

KU1 PSAR Section 1.3.2.1

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Preliminary barrier performance criteria (leakage) established based on meeting dose criteria for postulated LBEs with release, in accordance with SECY-18-0096 Enclosure 2.

Capability targets for radionuclide retention barriers initially established using preliminary leakage rates from CP analyses which satisfy dose acceptance criteria.

Design and analysis iterations continue post-CP to develop design and analysis margins. Ensure the final design satisfies leakage rate requirements set by the capability targets.

OL stage confirmatory evaluations:

Demonstrate acceptable doses using design leakage rates.

Demonstrate barrier can withstand LBE conditions.

Performance and Design Criteria 28 Leak rates used in bounding dose analysis Capability targets established based on leak rates Design and analysis iterations Update capability targets Confirmatory evaluations demonstrate adequate barrier performance

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Functional Containment Overview 29 Refueling Power Operations Power Operation Configuration Refueling Configuration

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Functional Containment Overview 30 Refueling Power Operations Primary Functional Containment Boundary & Performance Criteria Power Operation Configuration Major SSCs: reactor vessel, reactor head, RPA, PSPs, IVTM, CRD housing, IHX tubes/pipe wall, SPS-P piping up to isolation valve, SCG piping up to isolation valve, SCG isolation, seals, welds Refueling Configuration Major SSCs: Power Operation SSCs plus FTL, CATT, FFV, EVHM Leakage rate is set to ensure RSF performance criteria is met across core release DBAs and DBEs.

Activated SCG releases do not challenge DBA limits or F-C target curve - no additional HAA RR performance.

Activated SPS-P releases mitigated by hermetically sealed guard pipe - no additional HAA RR performance.

Intermediate sodium activity is below radionuclide screening threshold.

Head Access Area (HAA) Barrier Performance Functional Containment Overview 31 Event/Source PFCB HAA Barrier Representative Dose (LBE ID)

Dose Criteria In-vessel failed fuel event (DBA)

Intact and isolated (1% cover gas vol per day)

Not credited (100% HAA vol per day)*

0.96 rem 30-day TEDE (LFF-SAO-CN) 25 rem 30-day TEDE and worst 2-hr period (10 CFR 50.34)

In-vessel failed fuel event (1 assembly)

(DBE)

Intact and isolated (1% cover gas vol per day)

Not credited (100% HAA vol per day)*

0.237 rem 30-day TEDE (LFF-SAO-1)

Target 1 rem 30-day TEDE (F-C target curve)

In-vessel failed fuel event (2/3 core)

(BDBE)

Intact and isolated (1% cover gas vol per day)

Isolated (10% HAA vol per day) 0.406 rem 30-day TEDE (DHP-LOOP-3)

Target 25 rem 30-day TEDE (F-C target curve)

(*) Post-CP stage analyses will confirm whether the results are sensitive to this barrier performance and ensure it is adequately classified and appropriately applied in analyses.

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 Source Term and Radiological Consequence Analysis

Topical report NAT-9392-A, Radiological Source Term Methodology Report, describes the development of a mechanistic source term evaluation model utilized for the CP stage analysis.

Topical report NAT-9391-A, Radiological Release Consequences Methodology Topical Report, describes the development of non-DBA LBE, DBA, and control room habitability rad consequence evaluation model utilized for the CP stage analysis.

Source Term and Radiological Consequence Analysis Methods 33

Preliminary MST/Radiological Consequences (RC) analysis assigns consequences to release category end states in Step 3 (independent of frequencies and includes Other Quantified Events).

2.

PRA develops LBEs including event descriptions and frequencies in Step 4.

3.

Non-DBA MST/RC analysis performed for LBEs in Step 7a.

Results plotted on Frequency-Consequence (F-C) target curve, and input into cumulative risk measures.

4.

Prescriptive DBA MST/RC analysis performed in Step 7d and results evaluated against 10 CFR 50.34 dose limits.

PRA and MST/RC Interfaces 34

Reference:

NEI 18-04 Rev. 1, Figure 3-2 Preliminary MST/RC for end states Evaluate non-DBA MST/RC Evaluate DBA MST/RC

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Source Term Analysis 35 Example MST and In-vessel Transport Path In-pin radionuclide distribution Plenum fractions based on historical sodium bonded metal fuel data NaI and CsI chemical bonds form Failed pin release (cladding barrier failure)

Plenum releases instantly Flashing/vaporization of radionuclides Bubble entrainment Sodium Pool Gasses bypass pool Large pool height for effective decontamination factor Some vapors and aerosols removed Vaporization at Pool Surface Cover Gas Region and Compartments Radionuclide Decay and Daughtering Deposition of Aerosols Leakage Generic Approach (CP stage)

Event specific inventories

Fuel inventories & fuel decay times

System fluid activation concentrations

Bounding release quantities

All fuel pins rather than pin-level releases

Large quantities of system leaks and large ARFs assumed

Limit removal mechanisms in pool and compartment

Minimum Decontamination Factor (DF)

Minimum aerosol deposition

Limited exercising of best-estimate MST models in DBEs/BDBEs:

Flashing of RNs and bubble entrainment

Mechanistic bubble model

Vaporization of RNs at pool surface

Aerosol deposition in cover gas and compartment Representative MST/RC releases used to bound applicable LBEs.

RRCAT code

/Q factors aligned with most limiting release times per RG 1.183 Rev. 1

/Q factors envelop site-specific values

No deposition or plume depletion

Dose components:

Inhalation

Submersion

Semi-infinite plume conservatively assumed Non-DBA LBEs

MACCS code

Weather sampled from EPRI URD data

Ramsdell-Fosmire nearfield model

Ground deposition and decay modeled during plume transit

Dose components

Inhalation

Submersion

Ground shine

Submersion DCFs corrected for finite plume

MST Nominal Case Nominal Inputs Single Iteration MST Uncertainty Case Select Uncertainty Input Iterations Rad Con Case Weather Sampling Iterations Max release of iterations Pass release from single iteration Rad Con Case Weather Sampling Iterations 95th percentile result 95th percentile Dose Mean Dose 5th percentile Dose Mean result 5th percentile result

MST Bounding Case Inputs biased Conservative Pass release from single iteration Rad Con Case Weather Sampling Iterations 95th percentile Dose Mean Dose 5th percentile Dose 95th percentile result Mean result 5th percentile result

Pass release from single iteration Rad Con DBA Case Enveloping /Q DBA Dose Pass release from single iteration MST Bounding Case Inputs biased Conservative

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved SAO event family and source term case is a generic one assembly release to bound in-vessel at-power releases.

All fuel pins of the effected assembly are assumed to fail.

SCG pathways to/from the reactor vessel head isolate to complete PFCB.

Minimal impact on hot pool temperature and PHT conditions.

Hot pool temperature remains near normal conditions LFF-SAO-CN Single Assembly Overheat (SAO) 40 Head Access Area HAA leakage (HAA to Environment)

PFCB intact leak rate = 1% Cover Gas volume per day (Cover Gas to HAA)

SCG Exhaust to RWG (isolated)

Sodium Pool Core Cover Gas Release of one high BU assembly The postulated initiating event is a blockage of fuel subchannels or other localized faults within the reactor core during full power operating conditions, resulting in fuel damage. (Section 3.9.6.1 KU1 PSAR)

Instantaneous release and transport path:

RNs from plenum bubble rise / pool scrubbing release to cover gas Cover gas region:

Aerosol deposition

Holdup and decay (i.e., PFCB leakage to HAA)

HAA compartment:

No credit for aerosol deposition

Minimal holdup and decay (i.e., HAA leakage to environment)

Release to environment:

DBA Consequence evaluation model used for offsite dose

Enveloping /Q factors LFF-SAO-CN (continued)

Single Assembly Overheat (SAO) 41 1% cover gas vol per day (unfiltered) 100% HAA vol per day*

(unfiltered)

Noble gases/vapor: bypass pool Aerosols/particulates: small pool DF

(*) Post-CP stage analyses will confirm whether the results are sensitive to this barrier performance and ensure it is adequately classified and appropriately applied in analyses.

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 Major Accident

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Major Accident Evaluation Methods 43 10 CFR 50.34(a)(1)(ii)(D) directs license applicants to analyze a major accident event sequence with a release into containment.

Major accidents are postulated to result in potential hazards not exceeded by those from any accident considered credible.

Major accidents have generally been assumed to result in substantial core meltdown with subsequent release of an appreciable amount of fission products.

RG 1.253 Rev. 0 offers guidance for non-LWRs to fulfill requirements of 50.34(a)(1)(ii)(D).

TerraPower has chosen Option 1 from paragraph C.3.b: Use DBA dose consequence results from an LMP-based approach to fulfill dose requirements.

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Major Accident Evaluation Methods 44 An event should be selected as the major accident. Justification for selection is necessary if applicant selects a DBA that does not meet the general assumptions of a major accident (maximum credible release, core damage)

LFF-SAO-CN (Local Fuel Fault) represents the maximum core damage and associated release for all in-vessel DBAs analyzed for Kemmerer Unit 1 Uncertainty analyses for the mechanistic source terms and radiological doses should be described as part of the evaluation of conservative assumptions used in DBA analysis

Conservative selection of MST inputs as described in NAT-9392-A

Conservative selection of RC inputs as described in NAT-9391-A

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Major Accident Evaluation Methods 45 Total calculated radiological consequences for the postulated fission product release meets the following reference values for public dose from 10 CFR 50.34(a)(1)(ii)(D):

An individual on EAB for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following release receives less than 25 rem TEDE

An individual at LPZ boundary exposed to the entire release period receives less than 25 rem TEDE

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 DID and Integrated Risk Evaluation

The release category end-states are assigned radiological consequences to support evaluation of the following:

30-day TEDE at the EAB

Average individual latent cancer risk within 10 miles of the EAB

Average individual early fatality risk within 1 mile of the EAB

Probability of 100 mrem 30-day TEDE at the site boundary

The radiological consequences are assigned prior to LBE classification

Includes other quantified events (OQEs)

Integrated Risk Evaluation 48

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.

The DID adequacy evaluation is focused on the plant capability DID at the CP stage.

Review risk-significant LBEs for margin adequacy

A systematic review is performed for each LBE family

Determines if a single design or operational feature is relied upon for all five layers of defense

Confirms sufficient independence exists between DLs

Confirms prevention-mitigation balance is maintained DID Adequacy Evaluation 52

The systematic review qualitatively assesses the DID adequacy across the Fundamental Safety Functions and the five layers of defense

Identifies credited and failed PSFs for all LBEs (including DBAs) in the LBE family

Assigns each LBE in the LBE family to the appropriate layer of defense

Reviews each PSF to determine if all five layers rely on success of that function

If a function is relied upon for every layer of defense, the next available function is noted DID Adequacy Evaluation 53

An integrated decision-making process (IDP) is established to develop, review, and approve the following:

LBE selection

SSC safety classification

Special treatment selection

DID adequacy evaluation

As part of IDP, an integrated decision-making process panel (IDPP) is established to use deterministic and probabilistic inputs to review and assess the adequacy of the DID.

IDP Overview 54

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 Overview of Plant Transients and LBEs

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved CP stage analyses are developed in an appropriately conservative manner Biasing of key inputs, modeling approaches, acceptance criteria Generally applied DBA conservatisms to all LBE categories Plant responses to transients are generally stylized and grouped based on similar characteristics. Example groupings:

Reactivity transient Rod withdrawal, seismic, overcooling Loss of primary flow Loss of offsite power (both PSPs/ISPs), single PSP trips, single PSP locked rotor Loss of heat removal Loss of Nuclear Island Salt System (NSS)/Energy Island, loss of 1 intermediate sodium loop Preliminary Analyses 57

Inadvertent continuous withdrawal of a single control rod during full power operating conditions

Reactor scrams on either high high power range neutron flux, high high power-to-flow ratio, or high high positive power range neutron flux rate

Rod withdrawal inhibit SSC fails

Safe shutdown is achieved with IAC non-passive mode providing decay heat removal and fuel performance acceptance criteria are met.

General modeling conservativisms are used for this event.

Reactivity Transient Overview 58

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved Beyond Design Basis Event: Loss of offsite power with gravity rod insertion failure (DHP-LOOP-3)

CRDM receive scram signal; rods fail to insert from gravity. CRD driveline scram follow occurs and drives rods in at motor speed

System thermal hydraulic analysis performed in SAS (SAS4A/SASSYS-1). Figures of merit fail screening criteria indicating detailed fuel performance analysis necessary

Fuel performance code determined 2/3 core failure (fresh/once burned fuel doesnt fail)

Analysis performed with best-estimate plus uncertainty approach; however, several conservatisms are still retained for this event.

Loss of Primary Flow Overview 61

Loss of heat removal from one intermediate sodium loop Modeled as instantaneous loss of heat transfer to one IHX Reactor scram on high high cold pool temperature PSPs and ISPs trip on high high cold pool temperature, and the PSPs coast down Thermal capacity of the piping, components, and intermediate sodium in the remaining intermediate sodium loop are assumed to be available On scram, NSS isolated (SHX heat removal lost for intact intermediate sodium loop) and parasitic IAC heat removal lost Safe shutdown is achieved with RAC providing decay heat removal and fuel performance acceptance criteria are met.

General modeling conservativisms are used for this event.

Loss of Heat Removal Overview 64

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2025 TerraPower, LLC. All Rights Reserved ANS - American Nuclear Society AOO - Anticipated Operational Occurrence ASME - American Society of Mechanical Engineers BDBE - Beyond Design Basis Event CATT - Core Assembly Transfer Tube CCF - Common Cause Failure CFR - Code of Federal Regulations CP - Construction Permit CPA - Construction Permit Application CRD - Control Rod Drive System CRDM - Control Rod Drive Mechanism DBA - Design Basis Accident DBE - Design Basis Event DID - Defense In Depth DF - Decontamination Factor DL - Defense Layer EAB - Exclusion Area Boundary EPRI - Electric Power Research Institute EPZ - Emergency Planning Zone EVHM - Ex-Vessel Handling Machine F-C - Frequency-Consequence FHE - Ex-Vessel Fuel Handling System FFV - Fueling Floor Valves HAA - Head Access Area IAC - Intermediate Air Cooling System IE - Initiating Event IDP - Integrated Decision-Making Process IDPP - Integrated Decision-Making Process Panel IHX - Intermediate Heat Exchanger ISP - Intermediate Sodium Pump KU1 - Kemmerer Unit 1 LBE - Licensing Basis Event LPZ - Low Population Zone LWR - Light Water Reactor MST - Mechanistic Source Term NEI - Nuclear Energy Institute NHV - Nuclear Island Heating, Ventilation, and Air Conditioning System NST - Non-Safety-Related with No Special Treatment NSRST - Non-Safety-Related with Special Treatment NSS - Nuclear Island Salt System OL - Operating License OQE - Other Quantified Event PFCB - Primary Functional Containment Boundary PIE - Postulated Initiating Event PIRT - Phenomena Identification and Ranking Tables POS - Plant Operating State PRA - Probabilistic Risk Assessment PSAR - Preliminary Safety Analysis Report PSF - PRA Safety Function PSP - Primary Sodium Pump QHO - Quantitative Health Objectives RAC - Reactor Air Cooling System RC - Radiological Consequences RES - Reactor Enclosure System RG - Regulatory Guide RR - Radionuclide Retention RXB - Reactor Building SCG - Sodium Cover Gas System SPS-P - Primary Sodium Processing System SR - Safety-Related SSC - Structures, Systems, and Components TEDE - Total Effective Dose Equivalent Acronyms 67

ML25282A002

Contents

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Navigation menu

ML25282A002

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Navigation menu

Search