ML22262A252

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Presentation - 09/20/2022 Scale Melcor Sodium Fast Reactor (Sfr) Workshop
ML22262A252
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Issue date: 09/20/2022
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DE-NA0003525 SAND2022-12412PE
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SAND2022-12412 PE SCALE/MELCOR Non-LWR Source Term Demonstration Project - Sodium Fast Reactor (SFR)

September 20, 2022 Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energys National Nuclear Security Administration under contract DE-NA0003525. SAND20XX-XXXX P

Outline NRC strategy for non-LWR source term analysis Project scope Overview of Sodium Fast Reactor (SFR)

SFR reactor fission product inventory/decay heat methods & results MELCOR SFR model SFR plant model and sample analysis Summary 2

Integrated Action Plan (IAP) for Advanced Reactors Near-Term Implementation Action Plan Strategy 1 Strategy 4 Knowledge, Skills, Industry Codes and Capacity and Standards Strategy 5 Strategy 2 Technology Analytical Tools Inclusive Issues ML17165A069 Strategy 3 Strategy 6 Flexible Review Communication Process 3

IAP Strategy 2 Volumes These Volumes outline the specific analytical tools to enable independent analysis of non-LWRs, gaps in code capabilities and data, V&V needs and code development tasks.

Introduction Volume 1 ML20030A174 ML20030A176 Volume 3 Volume 2 Volume 4 Volume 5 ML20030A177 ML21085A484 ML21088A047 ML20030A178 4

NRC strategy for non-LWR analysis (Volume 3) 5

Role of NRC severe accident codes 6

Project Scope Project objectives Understand severe accident behavior

  • Provide insights for regulatory guidance Facilitate dialogue on staffs approach for source term Demonstrate use of SCALE and MELCOR
  • Identify accident characteristics and uncertainties affecting source term
  • Develop publicly available input models for representative designs 8

Project scope Full-plant models and sample calculations for representative non-LWRs 2021

  • Heat pipe reactor - INL Design A
  • Pebble-bed gas-cooled reactor - PBMR-400
  • Pebble-bed molten-salt-cooled - UCB Mark 1
  • Public workshop videos, slides, reports at advanced reactor source term webpage 2022
  • Molten-salt-fueled reactor - MSRE - public workshop 9/13/2022
  • Sodium-cooled fast reactor - ABTR - public workshop 9/20/2022 2023
  • Additional code enhancements, sample calculations, and sensitivity studies 9

Project approach

1. Build SCALE core model and MELCOR full-plant model
2. Select scenarios that demonstrate code capabilities
3. Perform simulations
  • Use SCALE to model decay heat, core radionuclide inventory, and reactivity feedback
  • Use MELCOR to model accident progression and source term
  • Perform sensitivity cases 10

Sodium Fast Reactor (SFR)

(US History)

Sodium fast reactors (1/4)

Experimental Breeder Reactor 1

  • Object to prove Enrico Fermis fuel breeding principle and to generate electricity Construction started in 1949 Uranium metal plate fueled with liquid sodium-potassium (NaK) coolant 1.4 MWth (200 kWe)and generated enough electricity to run four 200-W light bulbs (Dec 1951)

EBR Unit 1

rst_four_nuclear_lit_bulbs.jpeg

Experimental Breeder Reactor 2

  • Demonstrate a complete breeder reactor nuclear power plant Construction started in 1964 and reached full power in 1969 - 62.5 MWth and 20 MWe
  • Used uranium metal fuel rods Fuel designs researched and refined 96,399 uranium metal fuel slugs fabricated with 35,000 irradiated
  • Demonstrated passive safety tests Unprotected loss of flow Unprotected loss of heat removal EBR Unit II

[1]

12

Sodium fast reactors (2/4)

Fast Flux Test Facility (FFTF) reactor

  • Design started late 1970s and first criticality in 1982 400 MWth power rating National testing and research facility for advanced nuclear fuels, material, component, and passive safety features Generated medical isotopes and tritium for the US fusion program Overlapped EBR-2 and operated for ~10 years
  • Used mixed oxide metal fuel design Same construction as EBR-2 40,000 fuel pins irradiated with only one fuel pin cladding failure Goal burn-up to 100 GWd/MTM (achieved maximum burn-up of 238 GWd/MTM)
  • Important testbed for instrumentation and safety features Instrumentation to verify natural circulation Guard vessel around the reactor vessel to contain sodium spills FFTF Reactor Core and Vessel

[FFTF-20083, Rev 0]

13

Sodium fast reactors (3/4)

Fermi 1 - Prototype breeder reactor (200 MWth and 68 MWe)

  • Construction began in 1956 & operated from 1963 to 1972
  • Exhibited coolant flow blockage on October 5, 1966 Zr plate, near the bottom of the reactor, became loose and blocked the inlet nozzles - restricted sodium coolant flow 2 damaged fuel assemblies - resulting in partial fuel melts No radionuclide release to the environment but Fermi 1 underwent an extended shutdown for clean-up and repairs Restarted and ran from 1970 to 1972 Fermi Unit 1

[2]

14

Sodium fast reactors (4/4)

Clinch River Liquid Metal Fast Breeder Reactor Project

  • Authorized in 1970 1000 MWth and 350 MWe Mixed oxide (plutonium & uranium) fuel in 108 fuel assemblies
  • Stimulated advances in research, design, component fabrication, safety analysis, and licensing Fabrication of $380M of major components delivered (~50% of planned components)
  • Licensing activities started in 1974 Environmental Impact Statement approved in 1977 ASLB issued memorandum of findings in 1984 that all issues related to the construction permit had been addressed 250,000 pages of documentation for the licensing effort
  • Project terminated in 1983 DOE concluded the project demonstrated the ability to license LMBRs Clinch River Project [3] 15

ABTR - Reactor Design

  • Selected for the SCALE/MELCOR SFR demonstration
  • ABTR Design Specifics 250 MWth Pool-type SFR, near atmospheric pressures 355 core inlet / 510 core outlet 1260 kg/s core flowrate 2 mechanical or EM pumps 2 internal intermediate heat exchangers
  • Design features Guard vessel Short-term fuel storage in the reactor Primary connects to an intermediate loop inside the vessel ABTR Vessel

[ANL-AFCI-173]

16

ABTR core

  • 199 hex assemblies HT-9 steel duct surrounds each assembly Small interstitial gap region between assemblies
  • Multiple assembly type and region core 24 inner core driver assemblies 30 outer core driver assemblies 6 fuel test locations 10 control assemblies (B4C) 3 material test assemblies 78 reflector assemblies (HT-9 pins) 48 shield assemblies (B4C)
  • Color coding identifies diverse functions and assembly materials ABTR Vessel

[ANL-AFCI-173]

17

ABTR fuel

  • A hex HT-9 alloy duct surrounds 217 fuel rods HT-9 cladding (melts at 1687 K)

Steel wire used to maintain spacing U-TRU-Zr10% metallic fuel 1.2 m argon gas plenum to accommodate expansion and fission gases ABTR Fuel

[ANL-AFCI-173]

18

SCALE SFR Inventory, Decay Heat, Power, and Reactivity Methods and Results

NRC SCALE/MELCOR Non-LWR Demonstration Project Objectives:

  • Develop approach and models for SCALE analysis to obtain:
  • Radionuclide inventory
  • System decay heat
  • Power profiles
  • Reactivity coefficients Challenges:
  • Full core depletion calculation
  • Fast neutron spectrum Approach:
  • Develop fully heterogeneous 3D model
  • Perform depletion of one cycle
  • Evaluate neutronic characteristics
  • Verify SCALE results with results in the open literature

[1] Y. I. Chang, et al. (2006). Advanced Burner Test Reactor Preconceptual Design Report, Technical Report ANL-AFCI 173, Argonne National Laboratory, Argonne, IL. ABTR Cross Section [1] 20

Workflow SCALE specific Generic End-user specific Other SCALE Inventory Binary Output Interface File MACCS Input SCALE Power SCALE Text Distributions MELCOR Input Output Kinetics Data

  • SCALE capabilities used
  • Codes:
  • Sequences:

KENO-VI 3D Monte Carlo transport CSAS for reactivity (e.g., control assembly ORIGEN for depletion worth)

TRITON for reactor physics & depletion

  • Data:

ENDF/B-VII.1 nuclear data library 21

ABTR Neutronics Summary

  • 250 MWth rated power
  • Fuel: U/TRU-10Zr (16.5-20.7% TRU
  • 4-months operating cycle content)
  • Cladding: HT-9 cladding
  • Fast spectrum for burning actinides
  • 4.05 tHM initial core loading
  • Reflector: HT-9 reflector assemblies
  • Absorber: B4C shield and control assemblies 340 cm core height

~85 cm fuel height

[2] T. K. Kim, Benchmark Specification of Advanced Burner Test Reactor, 3D SCALE Model ANL-NSE-20/65, Argonne National Laboratory, 2006. doi:10.2172/1761066. 22

SCALE Analysis Approach

  • Develop KENO model of the benchmark:
  • At hot conditions (considering radial and axial expansions as specified by the benchmark)
  • With Beginning of Equilibrium Cycle (BOEC) fuel
  • Simple model for criticality and discretized model for depletion calculation
  • Perform CSAS-KENO analysis with simple model at BOEC:
  • Verify eigenvalue (keff) and effective delayed neutron fraction (eff) by comparison with ANL ABTR design report [1] and INL publication [3]
  • Analyze 3D flux and fission rate profiles
  • Perform TRITON-KENO analysis with discretized model:
  • Deplete model for one cycle to obtain inventory at End of Equilibrium Cycle (EOEC)
  • Analyze reactivity and power profiles at EOEC
  • Provide inventories, power profiles, and reactivities to MELCOR
  • Perform additional sensitivity studies in support of MELCOR analysis

[1] Y. I. Chang, et al. (2006). Advanced Burner Test Reactor Preconceptual Design Report, Technical Report ANL-AFCI 173, Argonne National Laboratory, Argonne, IL.

[3] C. M. Mueller, et al. (2021). NRC Multiphysics Analysis Capability Deployment FY2021

- Part 2, Technical Report INL/EXT-21-62522, Idaho National Laboratory, Idaho Falls, ID. 23

SCALE Analysis Approach BOEC Fuel Material BOEC KENO Model Discretization KENO Model BOEC CSAS TRITON Depletion EOEC Fuel EOEC Criticality Calculation Calculation BOECEOEC Isotopics KENO Model BOEC BOEC Power Nuclide Tallies Reactivity keff ,eff Profiles Inventory EOEC CSAS Criticality Calculation MELCOR EOEC EOEC Xenon Reactivity keff ,eff Worth 24

SCALE Model Construction and Verification

Modeling Assumptions

  • Full-core 3D Monte Carlo with continuous energy physics
  • System state defined in ABTR benchmark specifications [2]
  • BOEC starting isotopics
  • Temperature at hot full power Fuel: 855K Structure: 735K Coolant: 705K Shield: 630K
  • Geometry considers thermal expansion of all components
  • Helium fill gas (assumed)
  • Minor assumptions were made for temperatures not explicitly defined in the benchmark Temperatures are given as a mix of material-specific and region-specific definitions 3D SCALE ABTR Core with Fission Density Overlay

[2] T. K. Kim, Benchmark Specification of Advanced Burner Test Reactor, ANL-NSE-20/65, Argonne National Laboratory, 2006. doi:10.2172/1761066. 26

ABTR Model Development

  • KENO 3D full core model built based on ABTR benchmark specifications
  • Barrel, as described with assemblies, was replaced with a cylindrical configuration
  • Examined 115 and 114.413 cm (expansion of barrel SCALE Model at coolant temperature)
  • Effect is statistically indistinguishable
  • Internal face of the barrel is coolant, while the external face of the barrel is void Note: The displayed SCALE model does not display coolant to avoid confusion with the withdrawn control assemblies.

ABTR Core [2]

27

Neutron Flux in the BOEC Core Upper Structure Upper Plenum

& Control Rods Active Core Lower Reflector Energy-dependent Flux Spectrum Lower Structure Total Flux Note: The displayed flux is the flux per fission neutron divided by the mesh voxel volume. (Linear Scale) 28

Verification of BOEC SCALE model

  • Verification of the BOEC* SCALE model was performed relative to:
  • ANL ABTR reference design description [1]
  • INL ABTR Multiphysics report [3]
  • EOEC values not available for verification

[1] Y. I. Chang, et al. (2006). Advanced Burner Test Reactor Preconceptual Design Report, Technical Report ANL-AFCI 173, Argonne National Laboratory, Argonne, IL.

[3] C. M. Mueller, et al. (2021). NRC Multiphysics Analysis Capability Deployment FY2021

- Part 2, Technical Report INL/EXT-21-62522, Idaho National Laboratory, Idaho Falls, ID. 29

Reactivity Effects Reactivity Coefficients

  • Litany of model perturbations were performed to calculate reactivity coefficients
  • Axial Fuel Expansion:
  • A 1% expansion was considered, representing a 575K increase in fuel temperature
  • Density was correspondingly adjusted
  • Radial Grid Plate Expansion:
  • Uniform, radial thermal expansion of the SS-316 grid plate (increasing assembly pitch)
  • Cold (293K) to operating (628K)
  • Pitch increase of 0.087 cm (0.6%)

31

Reactivity Coefficients, cont.

  • Fuel Density:
  • A 1% density reduction while conserving dimensions (decreasing mass)
  • Enhanced response relative to axial fuel expansion due to lost mass
  • Structure Density:
  • All HT-9 components (cladding, ducts, reflector, structure, followers, barrel)
  • A 1% density reduction results from a 720K increase (decreasing mass)
  • Flowing sodium was voided within fuel assembly ducts, active fuel region and above
  • Varied from literature values, but known issues exist in calculating void worth with homogenized methods common for SFRs, as well as an XS library dependence [4,5]

[4] W. S. Yang, et al. (2007).Preliminary Validation Studies of Existing Neutronics Analysis Tools for Advanced Burner Reactor Design Applications Technical Report ANL-AFCI-186, Argonne National Laboratory.

[5] NEA (2016).Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes Technical Report NEA/NSC/R(2015)9, Nuclear Energy Agency. 32

Reactivity Coefficients, cont.

  • Doppler:
  • Nine fuel temperatures were utilized to determine the Doppler coefficient
  • Logarithmic response expected from fast spectrum HPR experience, so coefficient is calculated as derivative at nominal fuel temperature (with respect to reactivity, not keff)
  • Linear approach can cause underestimation of Doppler coefficient

-0.079 cents/K linear with 2 points

-0.098 cents/K linear with 9 points 33

Reactivity Coefficients, cont.

  • Nine fuel temperatures were utilized to determine the Doppler coefficient
  • Logarithmic response expected from fast spectrum HPR experience, so coefficient is calculated as derivative at nominal fuel temperature (with respect to reactivity, not keff)
  • Linear approach can cause underestimation of Doppler coefficient

-0.059 cents/K linear with 2 points

-0.075 cents/K linear with 9 points 34

Fuel Depletion TRITON Modeling

  • CSAS-KENO input was converted to TRITON-KENO input for depletion
  • CSAS to TRITON conversion involves:
  • Fuel region discretization Individual assembly definitions (60 fuel assemblies) for radial discretization 10 axial zones per assembly for axial discretization 600 total depletion zones for power profiling and tracking inventory
  • Applying specific power for the system 61.7 MW/MTHM
  • Determining the appropriate number of depletion steps and spacing for accurate flux response evolution while maintaining computational efficiency 6 burnup points over the 4-month cycle
  • Nuclide tracking between depletion steps 95 relevant fission products and actinides
  • Depleting materials of interest (fuel)

All 600 discretized depletion zones 36

TRITON Modeling

  • Analysis of normalized axially integrated assembly power distribution informed grouping of assemblies:
  • Group 1 (0.7-0.9)
  • Group 2 (0.9-1.0)
  • Group 3 (1.0-1.1)
  • Group 4 (1.1-1.2)
  • Group 5 (1.2-1.3)
  • Grouping allows for simpler data transfer to MELCOR (5 radial groups, 10 axial zones) vs pointwise (600 depletion zones)

Power Map 37

TRITON Modeling

  • Analysis of normalized axially integrated assembly power distribution informed grouping of assemblies:
  • Group 1 (0.7-0.9)
  • Group 2 (0.9-1.0)
  • Group 3 (1.0-1.1)
  • Group 4 (1.1-1.2)
  • Group 5 (1.2-1.3)
  • Grouping allows for simpler data transfer to MELCOR (5 radial groups, 10 axial zones) vs pointwise (600 depletion zones)

ABTR Model with Color-Coded Assemblies 38

Power Distribution

  • Axial profile steady radially throughout the core
  • Upper regions are slightly more variable and lower power with control assemblies withdrawn and a lack of upper reflector
  • Axial profile provided as the resulting normalized power from all assemblies (Total)

Fuel and Control Assembly Cross Section [2] 39

EOEC Inventories and Decay Heat

  • A full-core, explicit assembly TRITON model was used to deplete from BOECEOEC, generating power and nuclide inventory distributions
  • Nuclide inventories are available for 600 depletion zones at 6 time points over the 4-month cycle
  • Information flow to MELCOR
  • OBIWAN utility from SCALE 6.3 converts ORIGEN binary concentration files into Inventory Interface JSON files (ii.json)
  • Python script converts ii.json to a MELCOR DCH input file (mass and Total decay heat after shutdown decay heat by element group) 40

Core Decay Heat after Shutdown

  • Top 10 decay heat-producing isotopes in the first day following shutdown
  • Inventory consistent with other reactor designs, except Tc-104 (T1/2 =18 min)
  • Tc-104 is a top 10 contributor to decay heat in the first 30 seconds (2.3%) and 30 days (3%)
  • Fission yield of Tc-104 ~10x higher for Pu-239 vs. U-235
  • Pu content of initial core is much higher than other designs
  • Notable for the difference from other designsnot magnitude 41

Core Decay Heat after Shutdown, cont.

  • Inventory in the first 30 days consistent with other reactor designs, except Cm-242
  • ~12% additional decay heat at

~100 days due Cm-242

  • Initial loading contains higher trans-uranic (TRU) concentrations
  • Cm-242 generated through Am-242 in activation chains Difference in Cm-242 contribution to decay heat between ABTR and PWR 42

Additional Studies in Support of MELCOR Analyses

Statistical Convergence of Power Distribution

  • In a Monte Carlo simulation, results have statistical errors
  • Random number seed variations allow an estimate of the average power and the corresponding statistical error
  • Estimating the error used here to confirm convergence
  • Max. error of 0.1%, average of 0.05%
  • Understanding the magnitude of the statistical error allows to distinguish impact of actual power perturbations from statistical noise Statistical error (%) in assembly power 44

Single Assembly Sodium Voiding Effect on Power Distribution

  • MELCOR scenario considers single assembly blockage
  • Specifics of scenario to be detailed by the MELCOR team
  • Effect of single assembly voiding was investigated to confirm that the provided nominal power profile is applicable
  • Comparison of power maps shows that most differences are at the level of statistical noise (<0.1%)
  • Blocked assembly shows 0.6%

difference in power Blocked assembly 45

Additional Worth Estimates

  • Control assembly worths were calculated at BOEC by calculating reactivity differences with insertion
  • Each bank is individually sufficient for subcriticality
  • Demonstrated agreement with the design report [1]
  • Xe-135 worth: 9 +/- 5 pcm (confirmed negligible for ABTR)

[1] Y. I. Chang, et al. (2006). Advanced Burner Test Reactor Preconceptual Design Report, Technical Report ANL-AFCI 173, Argonne National Laboratory, Argonne, IL. 46

Summary and Conclusions SCALE SFR Summary

  • Fast-spectrum SFR modeling with SCALE
  • Continuous energy Monte Carlo neutronics with KENO and ORIGEN for depletion are high-fidelity, system-independent
  • Consistency with code-to-code comparisons in all verification studies
  • Key results
  • Reactivity Coefficients (including non-linear Doppler)
  • Full 3D power distributions (axial profile is sufficient for MELCOR)
  • Inventory and Decay Heat Cm-242 and Tc-104 have notable (but small) differences in SFRs utilizing U/TRU-Zr10% compared to PWRs
  • Future SFR work:
  • Additional reactivity analyses for further insights into SFR behavior
  • Analyses of scenarios in the SFR fuel cycle (Volume 5) 3D SCALE ABTR Core with Fission Density Overlay 48

MELCOR Sodium Fast Reactor Models

Evolution of SFR Modeling 50

Modeling SFR Accidents with MELCOR MELCOR SFR Modeling

  • SFR materials
  • U-10Zr metallic fuel, HT-9 cladding, and sodium bond
  • Fast reactor point kinetics
  • Establishing initial conditions Decay heat, radionuclide inventory, and power distribution specification (SCALE)

Initial fission product gas distribution (gas plenum, closed and open pores)

Fuel expansion and swelling geometry

  • Core damage progression
  • Fuel melting
  • Clad pressure boundary failure, melting and candling
  • Degraded fuel region molten and particulate debris behavior
  • Radionuclide release and transport Gap and plenum release Molten fuel fission gas release Thermal release models 51

Fuel damage progression and radionuclide release Models added to simulate unique metal fuel behavior

  • Fuel melting prior to cladding failure
  • Evolution of closed pores to interconnected, open pores
  • Existing models of candling, molten pools, particulate debris Fission product release characterized by distinct phases
  • In-pin release - migration of fission products to fission product plenum and sodium bond
  • Gap release - burst release of plenum gases and fission products in the bond
  • Pin failure & release - radionuclide releases from hot fuel debris 52

Sodium Fire Modeling Atmospheric chemistry + aerosol generation

  • Implementation and validation of MELCOR o Spray model is based on NACOM spray model from BNL o Pool fire model is based on SOFIRE-II code from ANL
  • Ongoing benchmarks with JAEA F7 pool and spray fire experiments
  • Previous benchmarks to ABCOVE AB5 and AB1 tests Figure adapted from ANL-ART-3 53

GRTR - Generalized Radionuclide Transport and Retention Tracks fission products and determines how much is released from liquid to atmosphere Characterizes evolution of fission products between different physico-chemical forms GRTR mass transport modeling essential for understanding effect of sodium on source term

  • Retention in sodium of many important radionuclides as a function of solubility and vapor pressure
  • Bubble transport and bursting
  • Deposition on structural surfaces in sodium pool and core
  • Jet breakup and splashing 54

GRTR and Integral MELCOR Simulations GRTR Physico- Advective and Inputs to GRTR Chemical Transport Fission/Transmutation Model Dynamics Dynamics Radionuclide mass in (or released to) liquid pool Soluble radionuclide form mass Advection of radionuclides in liquid pool or atmosphere Chemical speciation Colloidal radionuclide form mass Pressure in hydrodynamic volume Deposited radionuclide form mass Temperature in regions of hydrodynamic volume (e.g., liquid and Decay of radionuclides in atmosphere) hydrodynamic control volume Coupling with ORIGEN Advective flows of liquid and atmosphere between hydrodynamic Gaseous radionuclide mass volumes 55

MELCOR SFR Plant Model and Source Term Analysis

Core Core nodalization - light blue lines

  • Subdivided into 15 axial levels and 8 radial rings
  • Core divided according to assembly power and function (similar to SFP modeling)

Ring 1 through 6 = 60 fueled assemblies combined according to power Ring 7 = 10 control and 3 material test assemblies Ring 8 = 78 reflector and 58 shield assemblies The 8 rings share a common inlet plenum and the lower cold pool Fluid flow nodalization - black boxes

  • Sodium enters through the inlet plenum and flows into the assemblies 57

MELCOR core region mapping to SCALE MELCOR Ring 8 (78 reflector SCALE and 58 shield assemblies)

Radial Zones 1

2 3

MELCOR Ring 7 (10 control 4 and 3 material test assemblies) 5 MELCOR radial mapping to SCALE MELCOR axial mapping SCALE Radial Zone (r) 1 2 3 4 5

  • 1 SCALE level per MELCOR Radial Zone (r) 6 5 4 3 2 1 MELCOR COR level Number of Assemblies 15 12 21 6 5 1
  • 2 SCALE levels per Assembly Power Factor 0.80 0.95 1.05 1.17 1.27 MELCOR CVH level 58

Vessel All primary system sodium is contained within the vessel Sodium exits into a hot pool and circulates through the shell side of 2 intermediate heat exchangers (iHX)

A redan (wall) separates the hot pool from the cold pool 2 EM or mechanical pumps circulate sodium into the vessel inlet Free surfaces at the top of the hot and cold pools Argon gas above the free surfaces with connection to the cover-gas system

  • Assumed leak path for fission products 59

Direct Reactor Auxiliary Cooling System (DRACS) 4 trains - 625 kW/train

  • 0.25% of rated power per train (passive mode)
  • Passive or forced circulation operation (only passive mode modeled)

Each train has 3 loops in series

  • Cold pool primary coolant circulates through DRACS heat exchanger
  • A Na-K secondary side loop transfers heat from the DRACS Damper min HX to the natural draft heat exchanger (NDHX) area is 1%

Pump-driven or passive (only passive flow modeled)

  • Air flows through the NDHX to the plant stack Fan-driven or passive (only passive flow modeled)

Start-up: Damper on air flow springs open 60

Containment Containment dome

  • Defense in depth feature - radiological release and external challenges Nitrogen-inerted guard vessel surround the reactor vessel
  • Contains sodium leak and maintains sodium level above the fuel Reactor cavity and air gap (i.e., not a safety system)
  • Forced air cooling of concrete Argon cover-gas above the reactor hot and cold pool regions
  • System piping is not specified in the design description
  • Assumed to be the source of radionuclide leakage Leak rate is consistent with LWR containments
  • 0.1% vol/day at 10 psig (design pressure)
  • Dome = 5,580 m3 61

MELCOR model inputs Equilibrium inventory and decay heat from SCALE Radial and axial power profiles from SCALE Reactivity feedbacks from ANL ABTR report [ANL-AFCI-173]

U-10Zr fuel properties from INL [INL/JOU-17-44020]

HT-9 cladding and duct properties from [Leibowitz] & Bison [Hales]

62

Scenarios Unprotected transient over-power (UTOP)

  • Failure of the control rods to insert Unprotected loss-of-flow (ULOF)
  • Trip of primary and intermediate sodium pumps
  • Failure of the control rods to insert Single blocked assembly
  • Single assembly blocked
  • Leak from the cover gas piping into the containment 63

Unprotected transient over-power (UTOP)

Initial and boundary conditions

  • Highest worth control rod (0.9$) withdraws over 51 sec at mechanically-limited rate
  • Primary and intermediate pumps continue to operate
  • Intermediate heat exchanger remains operating Sensitivity analysis on additional reactivity addition
  • Additional sensitivity calculations at 1.5$, 2.0$, & 2.5$, and 3.0$
  • Sensitivity calculations on intermediate loop heat removal (i.e., limited to

~280 MW or unlimited) 64

UTOP - Withdraw of highest-worth CR

  • The highest-worth CR withdraws over
  • The core power rises to 346 MW in response to 51 sec to insert 0.9$. the reactivity insertion but subsequently drops in response due to the strong negative fuel feedback.
  • The net reactivity initially increases but is subsequently balanced by the negative
  • The long-term power stabilizes at 280 MW feedbacks after the CR is withdrawn
  • The maximum intermediate loop heat removal was assumed to be limited to (280 MW) ~112% of rated Reactivity Feedbacks Total core and fission power Axial+radial expansion 1.0 400 U-Zr density U-Zr Doppler Core power 0.8 Na void 350 Fission Power Na density 0.6 CRs in 300 CRs out 0.4 Total 250 Feedback ($) Power (MW) 0.2 200 0.0 150

-0.2 100

-0.4 50

-0.6 0 1 10 100 1000 10000 1 10 100 1000 10000 Time (sec) Time (sec) 65

UTOP - Withdraw of highest-worth CR

  • A 952 K peak fuel temperature occurs at Vessel Liquid Temperatures 1000 100 sec due to the CR withdraw and reactivity insertion 950
  • The reactivity feedback and the fuel temperature 900 adjust to match the secondary heat removal 850 Temperature (K)
  • The hot pool at the core exit has a ~64 K 800 Peak fuel temperature temperature rise, which increases the core 750 HP - Core outlet HP - Upper vessel inlet temperature 700 Core inlet
  • Large margin to U-10Zr fuel melting 650 (1623 K) 600 550 500 1 10 100 1000 10000 Time (sec) 66

UTOP - CR worth sensitivity

  • A larger reactivity insertion leads to
  • A larger reactivity insertion leads to corresponding successively higher peak fuel temperatures higher peak core powers
  • The peak fuel temperature response is
  • The long-term core power reflects the assumed capacity approaching the sodium saturation of the intermediate loop heat removal (~280 MW) temperature (~1215 K) in the 3.0$ case
  • The core inlet temperature increases with higher reactivity insertions Peak fuel temperature Core Power 1300 700 600 3.0$ insertion 1200 Core exit Tsat 2.5$ insertion 3.0$ insertion 2.0$ insertion 500 2.5$ insertion 1.5$ insertion 1100 2.0$ insertion Temperature (K) 0.9$ insertion Power (MW) 1.5$ insertion 400 1000 0.9$ insertion 300 900 200 800 100 700 0 1 10 100 1000 10000 1 10 100 1000 10000 Time (sec) Time (sec) 67

UTOP - Unlimited intermediate loop heat removal

  • The fuel temperature does not decrease
  • The core inlet temperature remains near the rated following the reactivity addition since the condition but the exit temperature and the control rods remain withdrawn corresponding core temperature rise settles to
  • The core inlet temperature remains offset the insertion of additional reactivity approximately constant in all cases
  • Higher core power higher fuel temperature higher intermediate loop heat removal requirements Peak fuel temperature Core Power 1300 700 1200 600 3.0$ insertion Core exit Tsat 2.5$ insertion 3.0$ insertion 2.0$ insertion 2.5$ insertion 500 1100 1.5$ insertion 2.0$ insertion Temperature (K) 0.9$ insertion 1.5$ insertion Power (MW) 400 1000 0.9$ insertion 300 900 200 800 100 700 0 1 10 100 1000 10000 1 10 100 1000 10000 Time (sec) Time (sec) 68

UTOP - Unlimited intermediate loop heat removal Limited Heat Removal Capacity (~112%) Unlimited Heat Removal Capacity 1050 1050 3.0 $ insertion 3.0 $ insertion 950 2.5$ insertion 950 2.5$ insertion 2.0$ insertion 2.0$ insertion 1.5$ insertion 1.5$ insertion 850 0.9$ insertion 850 0.9$ insertion Temperature (K) Temperature (K)

Core Outlet Temperature 750 750 650 650 550 550 450 450 1 10 100 1000 10000 1 10 100 1000 10000 Time (sec) Time (sec) 1050 1050 3.0 $ insertion 2.5$ insertion 950 950 2.0$ insertion 3.0 $ insertion 1.5$ insertion 2.5$ insertion 0.9$ insertion 2.0$ insertion 850 850 1.5$ insertion Core Inlet Temperature Temperature (K) Temperature (K) 0.9$ insertion 750 750 650 650 550 550 450 450 1 10 100 1000 10000 1 10 100 1000 10000 Time (sec) Time (sec) 69

Unprotected loss-of-flow (ULOF)

Initial and boundary conditions

  • Primary and intermediate pumps trip resulting in no secondary heat removal
  • 4 DRACS trains are available in passive mode Sensitivity analysis on DRACS availability
  • 0, 1, 2, and 3 DRACS trains available 70

ULOF The initial fuel heatup has strong negative expansion, The net reactivity oscillates near zero after fuel density, and fuel Doppler fuel feedbacks that 1000 sec greatly offsets the positive sodium density feedback that shuts down fission Reactivity Feedbacks Reactivity Feedbacks 0.2 Axial+radial expansion 0.10 U-Zr density U-Zr Doppler 0.0 Na void Na density

-0.2 CRs in Axial+radial expansion CRs out

-0.4 U-Zr density Total Feedback ($) Feedback ($)

U-Zr Doppler

-0.6 Na void 0.00 Na density CRs in

-0.8 CRs out Total

-1.0

-1.2

-1.4 -0.10 1 10 100 1000 10000 100000 10000 20000 30000 40000 50000 60000 70000 80000 90000 100000 Time (sec) Time (sec) 71

ULOF The long-term core power matches the The fuel and vessel liquid sodium temperatures DRACS heat removal rate after 20,000 sec quickly stabilize (5.6 hr)

The fission power is 1000 kW at 10,000 sec The natural circulation flow moves heat from and gradually increases to offset the decrease the core, through the iHXs to the cold pool, and in decay heat through the DRACS Core & fission power and DRACS heat removal Vessel pool and peak fuel temperatures 1.E+06 1000 Core power 950 Fission Power Peak fuel temperature 1.E+05 4xDRACS HP - Core outlet 900 HP - Upper CP - DRACS inlet Temperature (K) 850 CP - DRACS outlet 1.E+04 Power (kW) 800 1.E+03 750 700 1.E+02 Start of the primary-side DRACS heat exchanger flow 650 1.E+01 600 1 10 100 1000 10000 100000 0 20000 40000 60000 80000 100000 Time (sec) Time (sec) 72

ULOF - with variable DRACS sensitivity

  • Core power eventually converges on the DRACS 1xDRACS case shows a small heatup but other DRACS cases have similar responses heat removal rate
  • Thermal inertia of the DRACS and vessel mitigate
  • Dampers are normally 1% open heatups Expansion of sodium leads to hot to cold pool spill-over and eventually a filled vessel in 1% damper case DRACS heat removal and core power Peak fuel temperature 10000 1100 4xDRACS core power Pool level 4xDRACS Dampers at 1% open reaches top 1050 1xDRACS Hot pool of the vessel 3xDRACS core power overflow to 3xDRACS 2xDRACS cold pool 1000 3xDRACS 2xDRACS core power 4xDRACS 1000 2xDRACS Temperature (K) 1xDRACS core power 950 Power (kW) 1xDRACS Dampers at 1% open 900 850 100 800 750 10 700 1 10 100 1000 10000 100000 0 20000 40000 60000 80000 100000 Time (sec) Time (sec) 73

Single blocked fuel assembly Initial and boundary conditions

  • Inlet to a fuel assembly is blocked
  • Primary and intermediate pumps remain running
  • Control rods are assumed to insert after an off-gas high-radiation signal
  • The cover gas system leaks in the containment Assumed radionuclide release pathway 74

Single blocked fuel assembly 75

Single blocked fuel assembly

  • The fluid in the duct starts voiding within
  • The fuel cladding temperature responses (below) also 3 seconds indicate the fuel temperature response
  • The assembly sodium is boiled and expelled
  • The cladding temperature rise pauses while the fuel melts and then increases to the cladding melting within ~10 sec temperature
  • The cladding melts and collapses when the minimum thickness reaches a structural integrity limit Blocked assembly liquid sodium level Fuel cladding temperatures by axial level 2.5 1800 1600 2.0 Clad melting Fuel melting Active core level (m) 1400 Temperature (K)

Level 10 1.5 Level 9 1200 Level 8 Indicates collapse Level 7 1.0 Level 6 1000 Level 5 Level 4 Level 3 0.5 800 Level 2 Level 1 0.0 600 1 10 100 1 10 100 Time (sec) Time (sec) 76

Single blocked fuel assembly Molten debris is supported by solid debris Solid debris is supported by lower fuel Molten debris in inlet plenum 77

Single blocked fuel assembly Radionuclide release fraction from the fuel based

  • After the cladding failure, there is a on whole core inventory 1.E-01 prompt release of the plenum gas Fraction representing 100% of radionuclides in the blocked assembly inventory followed by thermal releases 1.E-02 Release from from the hot debris gas plenum and 100% release fuel voids
  • The analysis assumed blockage of a 1.E-03 Xe Cs high-powered center assembly with Ba Release fraction (-)

approximately 2.2% of the core 1.E-04 Iodine Te radionuclides Ru Mo

  • 97% of the noble gases 1.E-05 Ce

1.E-06 Cd Ag 1.E-07 1.E-08 1.E-09 1 10 100 1000 Time (sec) 78

Single blocked fuel assembly

  • Xe bubbles through the hot sodium pool above the core to the gas space.
  • Leakage rate through the failed off-gas line to the containment Assumed the sweep flow of 1 reactor gas space change per hour persisted during the transient Xe environmental release is very small due to the large containment volume and the low leak rate
  • The cesium and other radionuclides retained in the sodium 1.E-01 1.E-01 Xe radionuclide distribution Cs radionuclide distribution 1.E-02 1.E-02 1.E-03 Release fraction (-)

Release fraction (-)

1.E-03 1.E-04 100% release Cesium retained Released in-vessel 100% release 1.E-04 In-vessel 1.E-05 Released In-vessel Containment Environment 1.E-05 1.E-06 1.E-07 1.E-06 1 10 100 1000 10000 1 10 100 1000 10000 Time (sec)

Time (sec) 79

MELCOR Summary MELCOR SFR Summary

  • MELCOR capabilities were demonstrated New phenomenological modeling added to MELCOR for SFRs
  • Capabilities for a broad range of SFR accident scenarios (e.g., UTOP, ULOF)
  • Key physics considered Neutronics Liquid metal thermal hydraulics Core heat-up and degradation Fission product release
  • Future work Modeling improvements and enhancements Fuel cycle analysis (Volume 5) 81

Concluding remarks

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