ML22353A109

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Scale/Melcor non-LWR Source Term Demonstration Project - Sodium Fast Reactor (Sfr)
ML22353A109
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Issue date: 09/20/2022
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NRC/NRR/DANU, Oak Ridge, Sandia
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SAND2022-12412 PE
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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 SCALE/MELCOR Non-LWR Source Term Demonstration Project - Sodium Fast Reactor (SFR)

September 20, 2022 SAND2022-12412 PE

2 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 Outline

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

4 IAP Strategy 2 Volumes ML20030A177 ML20030A174 ML20030A176 ML20030A178 ML21085A484 Introduction Volume 1 Volume 2 Volume 3 Volume 4 Volume 5 ML21088A047 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.

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

6 Role of NRC severe accident codes

Project Scope

8 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 Project objectives

9 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

10

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 Project approach

Sodium Fast Reactor (SFR)

(US History)

12 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)

Sodium fast reactors (1/4)

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]

13 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 Sodium fast reactors (2/4)

FFTF Reactor Core and Vessel

[FFTF-20083, Rev 0]

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

Construction began in 1956 & operated from 1963 to 1972 Used uranium metal fuel (26% enriched U-235 fuel)

92 fuel assemblies surrounded by 548 fuel assemblies of depleted uranium 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 Sodium fast reactors (3/4)

Fermi Unit 1

[2]

15 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 Sodium fast reactors (4/4)

Clinch River Project [3]

16 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 - Reactor Design ABTR Vessel

[ANL-AFCI-173]

17

  • 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 core ABTR Vessel

[ANL-AFCI-173]

18

  • 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 ABTR Fuel

[ANL-AFCI-173]

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

20 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 NRC SCALE/MELCOR Non-LWR Demonstration Project

[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]

21 SCALE capabilities used Codes:

KENO-VI 3D Monte Carlo transport

ORIGEN for depletion Data:

ENDF/B-VII.1 nuclear data library Sequences:

CSAS for reactivity (e.g., control assembly worth)

TRITON for reactor physics & depletion Workflow Power Distributions Other MACCS Input MELCOR Input SCALE Binary Output Inventory Interface File SCALE Kinetics Data SCALE specific Generic End-user specific SCALE Text Output

22 250 MWth rated power 4-months operating cycle Fast spectrum for burning actinides 4.05 tHM initial core loading Fuel: U/TRU-10Zr (16.5-20.7% TRU content)

Cladding: HT-9 cladding Coolant: sodium Reflector: HT-9 reflector assemblies Absorber: B4C shield and control assemblies ABTR Neutronics Summary 340 cm core height 3D SCALE Model

~85 cm fuel height

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

23

  • 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 SCALE Analysis Approach

[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.

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

SCALE Model Construction and Verification

26

  • 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 Modeling Assumptions 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.

27

  • 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 at coolant temperature)
  • Effect is statistically indistinguishable
  • Internal face of the barrel is coolant, while the external face of the barrel is void ABTR Model Development ABTR Core [2]

SCALE Model Note: The displayed SCALE model does not display coolant to avoid confusion with the withdrawn control assemblies.

28 Neutron Flux in the BOEC Core Total Flux (Linear Scale)

Energy-dependent Flux Spectrum Note: The displayed flux is the flux per fission neutron divided by the mesh voxel volume.

Active Core Lower Reflector Lower Structure Upper Structure Upper Plenum

& Control Rods

29

  • Verification of the BOEC* SCALE model was performed relative to:

ANL ABTR reference design description [1]

INL ABTR Multiphysics report [3]

Verification of BOEC SCALE model

  • 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.

Reactivity Effects

31

  • 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%)

Reactivity Coefficients

32

  • 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]

Reactivity Coefficients, cont.

[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.

33

  • 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)

Reactivity Coefficients, cont.

Linear approach can cause underestimation of Doppler coefficient

-0.079 cents/K linear with 2 points

-0.098 cents/K linear with 9 points

34

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)

Reactivity Coefficients, cont.

Linear approach can cause underestimation of Doppler coefficient

-0.059 cents/K linear with 2 points

-0.075 cents/K linear with 9 points

Fuel Depletion

36

  • 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

ORIGEN tracks >2,000 nuclides

Based on ORIGEN results, TRITON updates densities of 95 relevant fission products and actinides for the neutron transport calculation Depleting materials of interest (fuel)

All 600 discretized depletion zones TRITON Modeling

37

  • 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)

TRITON Modeling Power Map

38

  • 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)

TRITON Modeling ABTR Model with Color-Coded Assemblies

39 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)

Power Distribution Fuel and Control Assembly Cross Section [2]

40

  • 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 decay heat by element group)

EOEC Inventories and Decay Heat Total decay heat after shutdown

41 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

42 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

Additional Studies in Support of MELCOR Analyses

44

  • 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 Convergence of Power Distribution Statistical error (%) in assembly power

45

  • 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 Single Assembly Sodium Voiding Effect on Power Distribution Blocked assembly

46

  • 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)

Additional Worth Estimates

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

Summary and Conclusions

48

  • 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)

SCALE SFR Summary 3D SCALE ABTR Core with Fission Density Overlay

MELCOR Sodium Fast Reactor Models

50 Evolution of SFR Modeling

51 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 Modeling SFR Accidents with MELCOR

52 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

53 Sodium Fire Modeling Figure adapted from ANL-ART-3 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

54 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 GRTR - Generalized Radionuclide Transport and Retention

55 GRTR and Integral MELCOR Simulations Inputs to GRTR Model Radionuclide mass in (or released to) liquid pool Chemical speciation Pressure in hydrodynamic volume Temperature in regions of hydrodynamic volume (e.g., liquid and atmosphere)

Advective flows of liquid and atmosphere between hydrodynamic volumes GRTR Physico-Chemical Transport Dynamics Soluble radionuclide form mass Colloidal radionuclide form mass Deposited radionuclide form mass Gaseous radionuclide mass Advective and Fission/Transmutation Dynamics Advection of radionuclides in liquid pool or atmosphere Decay of radionuclides in hydrodynamic control volume Coupling with ORIGEN

MELCOR SFR Plant Model and Source Term Analysis

57 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

58 MELCOR core region mapping to SCALE MELCOR axial mapping 1 SCALE level per MELCOR COR level 2 SCALE levels per MELCOR CVH level MELCOR radial mapping to SCALE MELCOR Ring 8 (78 reflector and 58 shield assemblies) and 3 material test assemblies)

MELCOR Ring 7 (10 control 1

2 3

4 5

SCALE Radial Zones SCALE Radial Zone (r) 1 2

3 4

5 MELCOR Radial Zone (r) 6 5

4 3

2 1

Number of Assemblies 15 12 21 6

5 1

Assembly Power Factor 0.80 0.95 1.05 1.17 1.27

59 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

60 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 HX to the natural draft heat exchanger (NDHX)

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 Damper min area is 1%

61 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

62 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]

63 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 Scenarios

64 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)

Unprotected transient over-power (UTOP)

65 UTOP - Withdraw of highest-worth CR The highest-worth CR withdraws over 51 sec to insert 0.9$.

The net reactivity initially increases but is subsequently balanced by the negative feedbacks after the CR is withdrawn The core power rises to 346 MW in response to the reactivity insertion but subsequently drops in response due to the strong negative fuel feedback.

The long-term power stabilizes at 280 MW The maximum intermediate loop heat removal was assumed to be limited to (280 MW) ~112% of rated 0

50 100 150 200 250 300 350 400 1

10 100 1000 10000 Power (MW)

Time (sec)

Core power Fission Power Reactivity Feedbacks Total core and fission power

-0.6

-0.4

-0.2 0.0 0.2 0.4 0.6 0.8 1.0 1

10 100 1000 10000 Feedback ($)

Time (sec)

Axial+radial expansion U-Zr density U-Zr Doppler Na void Na density CRs in CRs out Total

66 UTOP - Withdraw of highest-worth CR A 952 K peak fuel temperature occurs at 100 sec due to the CR withdraw and reactivity insertion The reactivity feedback and the fuel temperature adjust to match the secondary heat removal The hot pool at the core exit has a ~64 K temperature rise, which increases the core inlet temperature Large margin to U-10Zr fuel melting (1623 K) 500 550 600 650 700 750 800 850 900 950 1000 1

10 100 1000 10000 Temperature (K)

Time (sec)

Peak fuel temperature HP - Core outlet HP - Upper vessel Core inlet Vessel Liquid Temperatures

67 UTOP - CR worth sensitivity A larger reactivity insertion leads to successively higher peak fuel temperatures The peak fuel temperature response is approaching the sodium saturation temperature (~1215 K) in the 3.0$ case A larger reactivity insertion leads to corresponding higher peak core powers The long-term core power reflects the assumed capacity of the intermediate loop heat removal (~280 MW)

The core inlet temperature increases with higher reactivity insertions Peak fuel temperature Core Power 0

100 200 300 400 500 600 700 1

10 100 1000 10000 Power (MW)

Time (sec) 3.0$ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion 700 800 900 1000 1100 1200 1300 1

10 100 1000 10000 Temperature (K)

Time (sec)

Core exit Tsat 3.0$ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion

68 700 800 900 1000 1100 1200 1300 1

10 100 1000 10000 Temperature (K)

Time (sec)

Core exit Tsat 3.0$ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion 0

100 200 300 400 500 600 700 1

10 100 1000 10000 Power (MW)

Time (sec) 3.0$ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion UTOP - Unlimited intermediate loop heat removal The fuel temperature does not decrease following the reactivity addition since the control rods remain withdrawn The core inlet temperature remains approximately constant in all cases The core inlet temperature remains near the rated condition but the exit temperature and the corresponding core temperature rise settles to offset the insertion of additional reactivity Higher core power higher fuel temperature higher intermediate loop heat removal requirements Peak fuel temperature Core Power

69 Limited Heat Removal Capacity (~112%)

UTOP - Unlimited intermediate loop heat removal Unlimited Heat Removal Capacity 450 550 650 750 850 950 1050 1

10 100 1000 10000 Temperature (K)

Time (sec) 3.0 $ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion 450 550 650 750 850 950 1050 1

10 100 1000 10000 Temperature (K)

Time (sec) 3.0 $ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion Core Outlet Temperature Core Inlet Temperature 450 550 650 750 850 950 1050 1

10 100 1000 10000 Temperature (K)

Time (sec) 3.0 $ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion 450 550 650 750 850 950 1050 1

10 100 1000 10000 Temperature (K)

Time (sec) 3.0 $ insertion 2.5$ insertion 2.0$ insertion 1.5$ insertion 0.9$ insertion

70 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 Unprotected loss-of-flow (ULOF)

71

-0.10 0.00 0.10 10000 20000 30000 40000 50000 60000 70000 80000 90000 100000 Feedback ($)

Time (sec)

Axial+radial expansion U-Zr density U-Zr Doppler Na void Na density CRs in CRs out Total

-1.4

-1.2

-1.0

-0.8

-0.6

-0.4

-0.2 0.0 0.2 1

10 100 1000 10000 100000 Feedback ($)

Time (sec)

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

72 ULOF The long-term core power matches the DRACS heat removal rate after 20,000 sec (5.6 hr)

The fission power is 1000 kW at 10,000 sec and gradually increases to offset the decrease in decay heat The fuel and vessel liquid sodium temperatures quickly stabilize The natural circulation flow moves heat from the core, through the iHXs to the cold pool, and through the DRACS 600 650 700 750 800 850 900 950 1000 0

20000 40000 60000 80000 100000 Temperature (K)

Time (sec)

Peak fuel temperature HP - Core outlet HP - Upper CP - DRACS inlet CP - DRACS outlet Vessel pool and peak fuel temperatures Core & fission power and DRACS heat removal 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1

10 100 1000 10000 100000 Power (kW)

Time (sec)

Core power Fission Power 4xDRACS Start of the primary-side DRACS heat exchanger flow

73 700 750 800 850 900 950 1000 1050 1100 0

20000 40000 60000 80000 100000 Temperature (K)

Time (sec)

Dampers at 1% open 1xDRACS 2xDRACS 3xDRACS 4xDRACS Hot pool overflow to cold pool Pool level reaches top of the vessel Peak fuel temperature 10 100 1000 10000 1

10 100 1000 10000 100000 Power (kW)

Time (sec) 4xDRACS core power 4xDRACS 3xDRACS core power 3xDRACS 2xDRACS core power 2xDRACS 1xDRACS core power 1xDRACS Dampers at 1% open ULOF - with variable DRACS sensitivity

  • Core power eventually converges on the DRACS heat removal rate
  • Dampers are normally 1% open 1xDRACS case shows a small heatup but other DRACS cases have similar responses
  • Thermal inertia of the DRACS and vessel mitigate 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

74 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 Single blocked fuel assembly

75 Single blocked fuel assembly

76 600 800 1000 1200 1400 1600 1800 1

10 100 Temperature (K)

Time (sec)

Clad melting Fuel melting Level 10 Level 9 Level 8 Level 7 Level 6 Level 5 Level 4 Level 3 Level 2 Level 1 0.0 0.5 1.0 1.5 2.0 2.5 1

10 100 Active core level (m)

Time (sec)

Single blocked fuel assembly The fluid in the duct starts voiding within 3 seconds The assembly sodium is boiled and expelled within ~10 sec The fuel cladding temperature responses (below) also indicate the fuel temperature response The cladding temperature rise pauses while the fuel melts and then increases to the cladding melting temperature The cladding melts and collapses when the minimum thickness reaches a structural integrity limit Indicates collapse Blocked assembly liquid sodium level Fuel cladding temperatures by axial level

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

78 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1

10 100 1000 Release fraction (-)

Time (sec) 100% release Xe Cs Ba Iodine Te Ru Mo Ce La U

Cd Ag Single blocked fuel assembly After the cladding failure, there is a prompt release of the plenum gas inventory followed by thermal releases from the hot debris The analysis assumed blockage of a high-powered center assembly with approximately 2.2% of the core radionuclides 97% of the noble gases

~6% of iodine and cesium Fraction representing 100% of radionuclides in the blocked assembly Radionuclide release fraction from the fuel based on whole core inventory Release from gas plenum and fuel voids

79 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1

10 100 1000 10000 Release fraction (-)

Time (sec) 100% release Released In-vessel 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1

10 100 1000 10000 Release fraction (-)

Time (sec) 100% release Released In-vessel Containment Environment 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 Xe radionuclide distribution Cs radionuclide distribution Cesium retained in-vessel

MELCOR Summary

81

  • 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)

MELCOR SFR Summary

Concluding remarks

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