ML23261C396

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Scale - Melcor Non-LWR Fuel Cycle Demonstration Project for a Sodium Fast Reactor
ML23261C396
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Issue date: 09/20/2023
From: Lucas Kyriazidis
NRC/RES/DSA/FSTB
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Kyriazidis L
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SCALE & MELCOR non-LWR Fuel Cycle Demonstration Project Sodium Fast Reactors NRCs Volume 5 - Public Workshop #2 September 20, 2023 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Office of Nuclear Reactor Office of Nuclear Material Regulations Safety and Safeguards 1

Outline

  • NRC Strategy for non-LWRs Readiness
  • Project Scope
  • SFR Nuclear Fuel Cycle
  • Overview of the Simulated Accidents
  • Nuclide inventory, decay heat, and criticality calculations in SCALE
  • Sodium Fast Reactor Modeling using MELCOR
  • Summary & Closing Thoughts 2

NRCs Strategy for Preparing for non-LWRs

  • NRCs Readiness Strategy for Non-LWRs Volume #1

- Phase 1 - Vision & Strategy Systems Analysis

- Phase 2 - Implementation Action Plans Volume #5 Volume #2 Nuclear Fuel Fuel Cycle IAP Strategy #2 Performance

  • IAPs are planning tools that describe: Computer Codes and

- Required work, resources, and sequencing of work to achieve Tools readiness Volume #3

  • Strategy #2 - Computer Codes and Review Tools Volume #4 Licensing &

Source Term,

- Identifies computer code & development activities Dose Consequence

- Identifies key phenomena

- Assess available experimental data & needs 3

Whats in Volume 5?

What system(s) are we analyzing?

What code(s) are we using?

What are the key phenomena being considered?

Are there any gaps in modeling capabilities of the selected codes? How do we close these gaps?

What data do we have & what data do we need?

IAP Strategy 2 Volume 5 ML21088A047 4

LWR Nuclear Fuel Cycle Regulations for the Nuclear Fuel Cycle

  • Protects onsite workers, public and the environment against radiological and non-radiological hazards that arise from fuel cycle operations.
  • Radiation hazards
  • Radiological hazards
  • Non-radiological (i.e., chemical) hazards
  • Applicable Regulations

Project Scope - Non-LWR Fuel Cycle

  • Stages in scope for Volume 5 Enrichment Fuel Utilization Fresh Fuel UF6 Transportation Fuel Fabrication (including on-site spent UF6 enrichment Transportation fuel storage)
  • Stages out of scope for Volume 5 Uranium Mining & Milling
  • Not envisioned to change from current methods.

Power Production

  • Successfully completed and leveraged from the Volume 3 - Source Term & Consequence work Spent Fuel Off-site Storage & Transportation
  • Large amount of uncertainties for non-LWR concepts & lack of information Spent Fuel Final Disposal
  • Large amount of uncertainties for non-LWR concepts & lack of information 6

Codes Supporting non-LWR Nuclear Fuel Cycle Licensing

  • NRCs comprehensive neutronics package
  • NRCs comprehensive accident progression and
  • Nuclear data & cross-section processing source term code
  • Decay heat analyses
  • Characterizing and tracking accident
  • Criticality safety progression,
  • Radiation shielding
  • Performing transport and deposition of
  • Radionuclide inventory & depletion generation radionuclides throughout a facility,
  • Reactor core physics
  • Performing non-radiological accident
  • Sensitivity and uncertainty analyses progression 7

Project Approach Representative Initial and Boundary Conditions

  • Build representative fuel cycle designs leveraging the Volume 3 designs
  • Identify key scenarios and accidents exercising key phenomena & models Identify &

Address Code Simulating Accidents Modeling Gaps Assessment around Key Phenomena

  • Build representative SCALE & MELCOR models and evaluate Sensitivity Studies 8

Representative Fuel Cycle Designs

  • Completed 5 non-LWR fuel cycle designs for -
  • Heat Pipe Reactor (HPR)- INL Design A
  • High Temperature Gas Reactor (HTGR) - Pebble Bed Modular Reactor (PBMR)-400
  • Fluoride-Salt Cooled Hight Temperature Reactor (FHR) - University of California, Berkeley (UCB) Mark 1
  • Molten Salt Reactor (MSR) - Molten Salt Reactor Experiment (MSRE)
  • Sodium-Cooled Fast Reactor (SFR) - Advanced Burner Test Reactor (ABTR)
  • Identifies potential processes & methods, for example:
  • What shipping package could transport HALEU-enriched UF6? What are the hazards associated?
  • How is spent SFR fuel moved? What are the hazards associated?
  • How is fissile salt manufactured for MSRs? What are the various kinds of fissile salt that may be used? What are the hazards?

Prototypic Initial and Boundary Conditions for the SCALE &

MELCOR Analyses 9

Overview of the SFR fuel cycle F. Bostelmann 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

Overview Initial project effort was to identify hazards across the SFR fuel cycle

  • Determine details of the fuel cycle stage based on publicly available information
  • Use ABTR as basis for fuel assembly details and for SFR operation
  • Consider metallic SFR fuel
  • Identify potential hazards and accident scenarios for each stage of the fuel cycle
  • Identify accidents independently of their probability for occurrence
  • Select accident scenarios to demonstrate SCALE/MELCORs capabilities 11

SFR Fuel Cycle with Once-Through Fuel Scenario for this stage studied in this workshop 12

SFR Fuel Cycle with Reprocessed Fuel Scenario for this stage studied in this workshop 13

E1: Enrichment

  • Enrichment of UF6 up to 19.75 wt.% 235U [High Assay Low-Enriched Uranium (HALEU)]
  • US facilities for uranium enrichment using gas centrifuges
  • Louisiana Energy Services (Urenco USA) in Eunice, NM Currently the only active commercial process for enrichment of up to 5 wt.% 235U in the US
  • Centrus Energy Corp in Piketon, OH First U.S. facility licensed for HALEU production DOE program, started in 05/19, revised in 03/22 Phase 1 (~1 year): installation of HALEU cascade, demonstration of production of 20 kg UF6 HALEU Phase 2 (1 year): production of 900 kg UF6 HALEU Phase 3 (3 year): production of 900 kg UF6 HALEU/year Major hazards:
  • UF6 liquid and vapor leaks from damaged pipes or cylinders
  • Criticality due to unintended accumulation of enriched U 14

T1: Transportation of UF6 ORANO DN30-X package for up to 20 wt% 235U enrichment:

30B-X cylinder similar to 30B cylinder, but with criticality control system (internal absorber structure)

Permissible mass in DN30-X:

Package design Enrichment limit Permissible UF6 mass DN30-10 10 wt.% 235U 1460 kg DN30-20 20 wt.% 235U 1271 kg DN30-X package DN30-X protective structural packaging (PSP) unchanged to DN30: outer PSP acts as a shock absorber during drop tests and as thermal protection in fire tests Major hazards: 30B-X cylinder

  • Criticality due to water accidents and container drop Ref.: ORANO Safety Analysis Report for the DN30-X Package

R1: Reprocessing of Spent Nuclear Fuel

  • Reprocessing currently not pursued in the US, but only considered here to demonstrate code capabilities
  • Electrometallurgical treatment technology was originally proposed by ANL and already performed for EBR-II fuel
  • Electrometallurgical processing:
  • Complete set of operations to capture actinide elements from spent fuel and recycle them as fuel materials
  • Process:
  • Steel vessel with cadmium layer and electrolyte salt at 500°C
  • Chopped fuel is loaded into the anode basket
  • Actinides transport via electric current
  • Cathode deposits (U/Pu) are consolidated by melting and ready for to be used in fuel slug fabrication Schematic of the electrometallurgical treatment used for metallic fuel from the EBR-II Refs.:

Major hazards: [1] National Research Council. 2000. Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report. Washington, DC: The National Academies Press.

[2] Fredrickson, G. L, et al. 2022. History and status of spent fuel treatment at the INL

  • Release of radiological materials Fuel Conditioning Facility. Progress in Nuclear Energy 143, 104037, 2022.

[3] J.J. Laidler, et al.. Development of pyroprocessingtechnology. Progress in Nuclear Energy, 31(1):131-140, 1997. 16

F1: Fabrication of Metallic Fuel

  • Based on US experience of SFR fuel manufacturing (EBR-I, EBR-II, FFTF)
  • Reduction of enriched uranium to metal
  • Reduction of UF4 or uranium oxides by metals (Ca, Mg, Al, Ba)
  • Electrolytic reduction of uranium oxide
  • Alloying and casting to form the metallic slug
  • Most widely used: vacuum induction melting, alloying agent containing Pu and Zr
  • Machining and thermo-mechanical processing to form metallic fuel pellet Major hazards:
  • Release of hazardous or corrosive chemicals
  • Criticality from misfeeding or mishandling of fuel
  • Release of radiological materials from leaking containers Ref.: N.L. LaHaye, D.E. Burkes, Metal Fuel Fabrication Safety and Hazards - TO NRC-HQ-25 T-005, Non LWR LTD2, Pacific Northwest National Laboratory, PNNL-28622, 2019.

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F2: Fabrication of Fuel Assemblies

1. Fuel rod fabrication:
  • Cladding tube is loaded with sodium to facilitate bonding
  • Fuel slugs are loaded into the cladding tube
2. Fuel assembly manufacturing Major hazards:
  • Release of hazardous or corrosive chemicals/gases
  • Criticality from misfeeding or mishandling of fuel Ref.: D. E. Burkes, et al. A US Perspective on Fast Reactor Fuel
  • Release of radiological materials or sodium from rods Fabrication Technology and Experience Part 1: Metal Fuels and Assembly Design. Journal of Nuclear Materials, 389:458-469, 2009.

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T2: Transportation of Fresh Fuel Assemblies to Plant

  • SFR fuel have so far been transported in DOE-certified casks, but not in commercial size transportation packages
  • Possible candidates: ES-3100 (used for transporting test reactor fuel) or other Type B shipping container
  • ES-3100:
  • Certified for a variety of uranium bearing materials, including metals, with enrichments up to 100 wt.% 235U.
  • Loading limits determined from enrichment, material form, and presence of spacers
  • Container length might limit SFR fuel type to be transported Major hazards:
  • Criticality due to water accidents and container drop ES-3100
  • Reaction of sodium with water, air, or concrete in case of Ref.: J. Jarrell, A Proposed Path Forward for Transportation of container ruptures High-Assay Low-Enriched Uranium, INL Technical Report, INL/EXT-18-51518 Rev 0 (2018).

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U1/U2/U4 - Utilization Stages Ref.: Advanced Burner Test Reactor (ABTR)

  • Power: 250 MWt
  • Fuel: metallic U/TRU-Zr
  • Inner core assemblies:
  • 16.5% TRU fraction, 12 cycle lifetime, up to 94.5 GWd/tHM burnup
  • Outer core assemblies:
  • 20.7% TRU fraction, 15 cycles lifetime, up to 92.6 GWd/tHM burnup
  • Refueling for ~10 hours per assembly
  • Operation for cycle time of 4 months followed by refueling of a maximum of 7 components:
  • 2 inner, 2 outer, 0-1 test, 0-1 control Ref.: Y. I. Chang, et al., Advanced Burner Test Reactor Preconceptual Design Report, ANL-ABR-1/ANL-AFCI-173, Argonne National Laboratory, 2006.

20

U1/U2/U4: Major Components for Fuel Handling

  • Pantograph fuel handling machine and rotatable plug: Transfer of fuel assemblies into the core, within core and into a storage rack, and from the core
  • Storage rack: fresh and spent fuel assemblies, 36 positions Rotatable plug Pantograph
  • Fuel unloading machine: inserting and retrieving core assemblies from the cue position on the storage rack; heating, cooling and inert gas atmosphere for transferring fuel assemblies between the core and an IBC Storage
  • Intra-building casks (IBC): lead-shielded inter- rack building casks with inert gas atmosphere, with or without active cooling
  • Intra-building transfer tunnel: transfer of assemblies within inter-building cask Ref.: Y. I. Chang, et al., Advanced Burner Test Reactor Preconceptual Design Report, ANL-ABR-1/ANL-AFCI-173, Argonne National Laboratory, 2006.

21

U1/U2/U4: Major Hazards Major hazards:

  • Reaction of sodium with water, air, or concrete
  • Inadequate heat removal due to early removal of assembly from core or insufficient cooling by cask
  • Damage to fuel assembly causing fission product release
  • Criticality due to incorrect assembly pickup and drop off locations (consider sodium opaqueness)

Ref.: Y. I. Chang, et al., Advanced Burner Test Reactor Preconceptual Design Report, ANL-ABR-1/ANL-AFCI-173, Argonne National Laboratory, 2006.

22

Summary Major differences in the SFR fuel cycle compared to LWR:

  • Use of U-Zr (HALEU) fuel, U/TRU-Zr fuel, and potentially reprocessed fuel
  • No approved commercial size transportation and storage packages for SFR fuel assemblies with fresh fuel or reprocessed fuel
  • New chemicals and processes for metallic fuel fabrication
  • Remote fuel handling and high reliance on I&C due to opaqueness of sodium coolant Major identified hazards:
  • Higher enrichment impacting criticality during UF6 and fuel assembly storage and transportation
  • Hazards from the use of the various chemicals (spills, reaction with water, fire, explosion)
  • Sodium reaction with air and water, and sodium corrosion Additional details needed:
  • Fresh and spent fuel assembly storage details
  • Detailed SFR containment and building design
  • Details about specifications and operation of a reprocessing facility 23

Demonstration of SCALE for SFR Fuel Cycle Analysis D. Hartanto 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

OBJECTIVE AND APPLICATIONS Objective: Demonstrate use of SCALE for simulating accident scenarios in all stages of the nuclear fuel cycle for Sodium-cooled Fast Reactors (SFR)

Scenario 1: Release of fission products during operation / refueling (U3)

  • Accident: Seismic event causing the refueling machine to fall and release the fuel assembly.
  • Analysis: Determine fuel inventory and perform SCALE radiation dose calculations.

Scenario 2: Criticality event / fissile material buildup during reprocessing (R1)

  • Accident: Misfeed of material into the electro-processing batch leading to fissile material buildup / criticality as materials collect on the cathode.
  • Analysis: Determine fuel inventory and perform SCALE criticality calculations.

Scenario 3: Release of fission products during reprocessing (R1)

ABTR reactor building

  • Accident: A leak in the waste stream storage tank allows for release of fission products during reprocessing. Ref.: Chang, Y. I., et al. Advanced Burner Test Reactor -
  • Analysis: Determine fuel inventory and perform SCALE activity Preconceptual Design Report. Technical Report ANL-ABR-1 (ANL-AFCI-173), Argonne National Laboratory, 2006.

calculations.

25

OBJECTIVE AND APPLICATIONS Objective: Demonstrate use of SCALE for simulating accident scenarios in all stages of the nuclear fuel cycle for Sodium-cooled Fast Reactors (SFR)

Scenario 1: Release of fission products during operation / refueling (U3)

  • Accident: Seismic event causing the refueling machine to fall and release the fuel assembly.
  • Analysis: Determine fuel inventory and perform SCALE radiation dose calculations.

Scenario 2: Criticality event / fissile material buildup during reprocessing (R1)

  • Accident: Misfeed of material into the electro-processing batch leading to fissile material buildup / criticality as materials collect on the cathode.
  • Analysis: Determine fuel inventory and perform SCALE criticality calculations.

Scenario 3: Release of fission products during reprocessing (R1)

  • Accident: A leak in the waste stream storage tank allows for release of Electrorefiner fission products during reprocessing.
  • Analysis: Determine fuel inventory and perform SCALE activity Ref.: Pyroprocessing Technologies Brochure, calculations. Argonne National Laboratory 26

REFERENCE SODIUM FAST REACTOR DESIGN Advanced Burner Test Reactor (ABTR)

Reactor Power 250 MWt, 95 MWe Coolant Temperature 355°C/510°C Fuel Metallic Cladding and Duct HT-9 Cycle Length 4 months Inner FA Mid FA Outer FA Refs.:

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

[2] Kim, T. K. Benchmark Specification of Advanced Burner Test Reactor. Technical Report ANL/NSE-20/65, Argonne National Laboratory, 2020. 27

APPLIED SCALE6.3.1 SEQUENCES Rapid inventory generation Shielding & radiation dose Criticality calculation with with ORIGAMI calculations with MAVRIC CSAS

  • Depletion and decay solver
  • Monte Carlo photon and neutron
  • Monte Carlo neutron transport (ORIGEN) transport code (MONACO) with code (KENO or Shift) for criticality automated variance reduction for safety analysis
  • Requires pre-calculated ORIGEN shielding analyses cross-section libraries (generated
  • Output:

in previous work for the ABTR*)

  • Requires radiation source terms. - Multiplication factor

- Spatial flux and fission density

  • Output:
  • Output: distributions

- Nuclide inventory of irradiated fuel - Spatial flux/dose rate distributions

- Decay heat and activity of irradiated fuel

- Photon and neutron source terms of irradiated fuel

- Activation sources of irradiated non-fuel materials (Zr, HT9, and SS316)

Nuclide inventory and decay Ref:

[1] Wieselquist, W. A., Lefebvre, R. A., Eds., SCALE 6.3.1 User Manual, ORNL/TM-SCALE-6.3.1, Oak Ridge National Laboratory, heat of the irradiated fuel are 2023.

[2] *Shaw, A, et al. SCALE Modeling of the Sodium Cooled Fast-Spectrum Advanced Burner Test Reactor. Technical Report passed to MELCOR. ORNL/TM-2022/2758, Oak Ridge National Laboratory, 2022.

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SPENT NUCLEAR FUEL ABTR Source terms for all scenarios (ABTR TRU Inner)

TRU U/TRU-10Zr Fuels 16.5 wt.% (inner) & 20.7 wt.% TRU (outer) Gas Specific power: 65.6 GW/tHM (inner) & 51.4 GW/tHM (outer) plenum Discharged BU: 94.5 GWd/tHM (inner) & 92.6 GWd/tHM (outer)

ABTR Source terms for all scenarios HALEU U-10Zr Na bond Fuel 16.5 wt.% U-235 Specific power: 46.2 GW/tHM Discharged BU: 149.74 GWd/tHM Fuel ABTR PWR Source terms for scenarios 2 and 3 fuel assembly Fuel UO2 4.95 wt.% U-235 Specific power: 33.7 GW/tHM Discharged BU: 50.00 GWd/tHM Lower refl.

Refs.:

[1] Kim, T. K. Benchmark Specification of Advanced Burner Test Reactor. Technical Report ANL/NSE-20/65, Argonne National Laboratory, 2020.

[2] Natrium Clearpath Webinar (nationalacademies.org).

[3] Kim, T. K. and T. A. Taiwo, Fuel Cycle Analysis of Once-Through Nuclear Systems. Technical Report ANL-FCRD-308, Argonne National Laboratory, 2010.

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SPENT NUCLEAR FUEL Rapid inventory generation with ORIGAMI Irradiation history:

Gas

  • TRU Inner plenum
  • Loaded for 12 cycles
  • 120 days per cycle
  • TRU Outer Na bond
  • Loaded for 15 cycles
  • 120 days per cycle Fuel
  • Loaded for 6 cycles fuel assembly
  • 540 days per cycle
  • Assuming 10 days of cooling time between cycles Lower
  • Discharged fuel assembly is planned to be stored for 7 reactor cycles in the refl.

in-vessel storage (IVS) 30

SPENT NUCLEAR FUEL - COMPOSITION Composition distribution in the fuels at BOC and EOC (wt.%)

94.5 92.6 149.74 50.0 GWd/tHM GWd/tHM GWd/tHM GWd/tHM

  • Since all ABTR fuels have a higher burnup, they produce more TRUs and FPs than the PWRs.
  • More FPs are produced by ABTR HALEU fuel than U/TRU fuel due to higher burnup (~150 GWd/tHM).
  • ABTR U/TRU fuels have higher TRU fraction at EOC compared to the HALEU fuel.

BOC: beginning of cycle EOC: end of cycle FP: fission product TRU: transuranics 31

SPENT NUCLEAR FUEL - DECAY HEAT Top 5 decay heat contributors at 10 days and 5 years (ABTR) and *10 years (PWR)

At 10 days of At 5 years of Fuel cooling time cooling time U/TRU 140La (21%) 137mBa (22%)

106Rh (12%) 106Rh (14%)

Inner 144Pr (9%) 90Y (12%)

95Nb (8%) 238Pu (9%)

95Zr (8%) 134Cs (7%)

U/TRU 140La (21%) 137mBa (22%)

106Rh (12%) 106Rh (12%)

Outer 144Pr (9%) 90Y (12%)

95Nb (8%) 238Pu (10%)

95Zr (8%) 134Cs (6%)

HALEU 140La (21%) 90Y (29%)

144Pr (11%) 137mBa (29%)

95Nb (9%) 134Cs (11%)

95Zr (9%) 137Cs (7%)

106Rh (7%) 238Pu (6%)

PWR*

140La (21%) 90Y (25%)

144Pr (10%) 137mBa (25%)

  • Decay heat at shutdown is similar between the different fuel 95Nb (8%) 238Pu (11%)

types (~5-7% power) 106Rh (8%) 244Cm (11%)

  • Initially, slightly higher for the U/TRU inner fuel due to higher 95Zr (8%) 137Cs (7%)

specific power although its burnup is lower than HALEU 32

SPENT NUCLEAR FUEL - ACTIVITY Top 5 activity contributors at 10 days and 5 years (ABTR) and *10 years (PWR)

At 10 days of At 5 years of Fuel cooling time cooling time U/TRU 103Ru (8%) 137Cs (19%)

103mRh (8%) 137mBa (18%)

Inner 95Nb (7%) 241Pu (15%)

95Zr (7%) 147Pm (12%)

141Ce (6%) 90Y (7%)

U/TRU 103Ru (8%) 137Cs (19%)

103mRh (8%) 241Pu (19%)

Outer 95Nb (7%) 137mBa (18%)

95Zr (6%) 147Pm (11%)

141Ce (6%) 90Y (7%)

HALEU 95Nb (8%) 137Cs (22%)

95Zr (7%) 137mBa (21%)

103Ru (6%) 90Y (16%)

103mRh (6%) 90Sr (16%)

141Ce (6%) 147Pm (10%)

PWR*

140La (32%) 137mBa (76%)

95Nb (14%) 134Cs (16%)

  • Similar trends compared to decay heat 95Zr (13%) 154Eu (7%)
  • PWR has the lowest activity due to lower FPs built-up 103Ru (9%)

134Cs (6%)

125Sb (0.5%)

106Rh (0.2%)

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Scenario 1 Seismic event causing the refueling machine to fall and release the fuel assembly 34

CONTAINMENT BUILDING (CB) MODEL

  • Fuel assembly falls down from the refueling machine cask.
  • ABTR HALEU and U/TRU (Inner)
  • Case 1: Fuel assembly is cooled for 10 days
  • Case 2: Fuel assembly is cooled for 7 reactor cycles
  • 1.2-cm thick steel liner
  • Radiation dose rate inside and
  • Reinforced concrete outside of containment are (~1 m) assuming Fuel assembly calculated with MAVRIC using rebar-to-concrete intact fuel assembly as radiation mass ratio of 0.106 source (irradiated fuel and activation products).
  • ANSI standard (1977) flux-to-dose-rate factors
  • Cartesian and cylindrical mesh for
  • Fuel assembly dose calculations
  • Statistical error < 0.5%

Refs.:

[1] P. F. Peterson et al., Metal and Concrete Inputs for Several Nuclear Power MAVRIC model of the CB and 3D view of the CB with front Plants, Report UCBTH-05-001, 2005.

[2] Chang, Y. I., et al. Advanced Burner Test Reactor - Preconceptual Design unshielded fuel assembly quarter segment removed Report. Technical Report ANL-ABR-1 (ANL-AFCI-173), Argonne National Laboratory, 2006.

(front view) 35

NEUTRON SOURCE TERMS

  • Neutron sources from spontaneous fission Fuel light element impurities might contribute additional neutron sources Cooling time (d)

Cooling ABTR HALEU ABTR U/TRUHALEU time (inner) TRU 10 days Cm-242 (74.2%) Cm-242 (44.3%)

Pu-240 (17.3%) Cm-244 (54.0%)

7 cycles Pu-240 (71.3%) Cm-244 (94.1%)

Pu-238 (16.2%) Cm-242 (0.27%)

Cm-244 (11.5%)

Half life: 7 cycles of cooling time:

Cm-242: 162.8 d U/TRU: 840 d Cm-244: 18.10 y HALEU: 3780 d 36

GAMMA SOURCE TERMS

  • Strong fuel gamma radiation sources
  • Total dose rate dominated by fuel gamma dose rate
  • The neutron dose rate negligible as compared to the gamma dose rate (~6 orders of magnitude lower) 37

SENSITIVITY OF DOSE RATE TO FUEL ASSEMBLY LOCATION AND ORIENTATION Dose rate (mrem/h)

Position 1 ABTR HALEU Position 2 ABTR HALEU fuel 1 fuel assembly leaning on the assembly lying on the floor next 2 containment wall to containment wall (front view) (top view)

Location of the FA

  • Highest dose rate observed when fuel assembly leans on containment wall This model is used for all dose rate calculations 38

MAIN BETA AND GAMMA EMITTERS Nuclides important to the gamma source terms for both ABTR U/TRU and HALEU fuels

  • 10 days of cooling
  • 7 cycles of cooling Nuclide Half-life Nuclide Half-life Nuclide Half-life Y-91 58.5 d Cs-137/Ba-137m 30.07 yr/2.552 m Sr-90/Y-90 28.78 yr/2.67 d Zr-95 64.02 d Ba-140 12.75 d Ru-106/Rh-106 1.02 yr/2.18 h Nb-95 34.99 d La-140 1.678 d Ru-103 39.27 d Ce-144/Pr-144 284.6 d/17.28 m Ag-110m 249.8 d Ru-106/Rh-106 1.02 yr/2.18 h Nd-147 10.98 d Sb-125 2.758 yr Sb-124 60.2 d Pm-148m 42.3 d Cs-134 2.065 yr Te-132/I-132 3.2 d/2.28 h Eu-154 8.593 yr Cs-137/Ba-137m 30.07 yr/2.552 m Cs-134 2.065 yr Eu-156 15.2 d Cs-136/Ba-136m 13.16 d/0.308 s Ce-144/Pr-144 284.6 d/17.28 m Eu-152 13.54 yr Eu-154 8.593 yr 39

DOSE RATE MAP INSIDE CB 10 days cooling time Dose rate (rem/h) 4.6x106 rem/h 6.0x106 rem/h (4.6x104 Sv/h) (6.0x104 Sv/h) 7.0x102 rem/h 9.0x102 rem/h (7.0 Sv/h) (9.0 Sv/h)

ABTR HALEU ABTR U/TRU 40

DOSE RATE MAP INSIDE CB Dose rate (rem/h) 7 cycles of cooling time 9.2x104 rem/h (9.2x102 Sv/h) 1.9x105 rem/h 13.5 rem/h (1.9x103 Sv/h)

(0.135 Sv/h) 30 rem/h (0.3 Sv/h)

ABTR HALEU ABTR U/TRU 41

DOSE RATE MAPS OUTSIDE CB 10 days cooling time Dose rate (mrem/h) 530 mrem/h 720 mrem/h (5.3 mSv/h) (7.2 mSv/h) 0.4 mrem/h 0.5 mrem/h (4 Sv/h) Fuel (5 Sv/h) Fuel assembly assembly ABTR HALEU ABTR U/TRU 42

DOSE RATE MAPS OUTSIDE CB 7 cycles of cooling time Dose rate (mrem/h) 6.6 mrem/h 0.34 mrem/h (66 Sv/h)

(3.4 Sv/h) 0.2 rem/h 5 rem/h (2.0E-03 Sv/h) (5E-02 Sv/h)

Fuel Fuel assembly assembly ABTR HALEU ABTR U/TRU 43

COMPARISON OF MAXIMUM DOSE RATES Cooling time Inside CB Outside CB ABTR HALEU ABTR U/TRU ABTR HALEU ABTR U/TRU 10 days 4.6x106 rem/h 6.0x106 rem/h 530 mrem/h 720 mrem/h (4.6x104 Sv/h) (6.0x104 Sv/h) (5.3 mSv/h) (7.2 mSv/h) 7 cycles 9.2x104 rem/h 1.9x105 rem/h 0.34 mrem/h 6.6 mrem/h (9.2x102 Sv/h) (1.9x103 Sv/h) (3.4 Sv/h) (66 Sv/h)

  • For comparison, the irradiation dose of PWR spent fuel (50 GWd/tHM) after 10 days of cooling is about 1.7x106 rem/h (1.7x104 Sv/h).
  • Total dose rate dominated by primary gamma dose rate at these cooling times
  • 10 CFR 20.1201 occupational annual dose limit for adults Total effective dose equivalent (TEDE)* of 5 rems (0.05 Sv)
  • TEDE means the sum of the effective dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures) (10 CFR 20.1003).

44

Scenario 2 Misfeed of material into the electro-processing batch leading to fissile material buildup / criticality as materials collect on the cathode 45

ELECTROMETALLURGICAL PROCESSING

  • Electrometallurgical technology was originally proposed by ANL as a process to treat all DOE spent fuels.
  • The analyses in this work were based on the experience for EBR-II spent nuclear fuel treatment.
  • The chopped PWR spent fuel will undergo oxide reduction process (voloxidation) before electrorefining.
  • Fuel assemblies irradiation history:
  • ABTR U/TRU (Inner): 94.5 GWd/tHM + 5 years cooling
  • ABTR HALEU: 149.74 GWd/tHM + 5 years cooling
  • PWR: 50 GWd/tHM + 10 years cooling Ref.: National Research Council. 2000. Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report. Washington, DC: The National Academies Press. https://doi.org/10.17226/9883. 46

ELECTROREFINING Mark-IV Electrorefiner GBZ: Glass-bonded zeolite Ref.: National Research Council. 2000. Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report. Washington, DC: The National Academies Press. https://doi.org/10.17226/9883. 47

ELECTROREFINING GBZ: Glass-bonded zeolite Ref: Fredrickson, G. L, et al. 2022. History and status of spent fuel treatment at the INL Fuel Conditioning Facility. Progress in Nuclear Energy 143, 104037, 2022.

Ref.: National Research Council. 2000. Electrometallurgical Techniques for DOE Spent Fuel Treatment: Final Report. Washington, DC: The National Academies Press. https://doi.org/10.17226/9883. 48

CRITICALITY ANALYSIS OF ELECTROREFINER CSAS Model of Electrorefiner (40x40)

Single Cathode ER Dual Cathodes ER Anode basket Steel cathode Salt (12)

LiCl-KCl-PuCl3 (FP&TRU)

Cadmium pool (6)

Pure U (Dendritic)

Ref.: Robert D. Mariani, et al. Criticality Safety Strategy and Analysis Summary for the Fuel Cycle Facility Electrorefiner at Argonne National Laboratory West, Nuclear Technology, 114:2, 224-234, 1996. 49

CRITICALITY ANALYSIS OF ELECTROREFINER Vector of U and Pu in the recycled nuclear fuel Fuel U Vector Pu Vector U/TRU 0.01% U-234 0.08% U-235 0.53% Pu-238 78.45% Pu-239 0.03% U-236 18.81% Pu-240 99.88% U-238 1.47% Pu-241 0.74% Pu-242 HALEU 0.01% U-234 5.81% U-235 0.93% Pu-238 88.83% Pu-239 2.66% U-236 9.74% Pu-240 91.52% U-238 0.47% Pu-241 0.04% Pu-242 PWR 0.02% U-234 0.80% U-235 3.12% Pu-238 54.18% Pu-239 0.63% U-236 25.74% Pu-240 98.55% U-238 9.59% Pu-241 7.73% Pu-242 50

CRITICALITY ANALYSIS OF ELECTROREFINER Multiplication factor as function of U mass in cathode

  • keff is clearly below 0.95 even with the maximum U mass in the cathode
  • Similar results were obtained by both nuclear data libraries (ENDF/B-7.1 and ENDB/B-8.0)

Single Cathode ER Dual Cathodes ER 51

CRITICALITY ANALYSIS OF ELECTROREFINER Multiplication factor as function of salt height in the tank

  • keff does not change significantly when the salt height increases, and remains clearly below 0.95 Single Cathode ER Dual Cathodes ER 52

Scenario 3 A leak in the waste stream storage tank allows for release of fission products during reprocessing 53

ACTIVITY OF SALT

  • According to IAEA Technical Reports Series No. 135 (1972), an activity > 10-2 Ci/ml requires cooling and shielding.
  • Currently salt is assumed to contain 10 wt.% PuCl3. Pu in PuCl3 lumps all TRU and majority of the fission products.

54

Summary 55

Summary & Outlook

  • SCALE capabilities to simulate different scenarios in the different SFR fuel cycle stages were demonstrated.
  • The demonstrated capabilities included the rapid calculation of fuel inventory, decay heat and activity, as well as shielding, radiation dose, and criticality calculations.
  • Key observations:
  • The radiation dose of the ABTR spent fuel is significant, requiring proper shielded when removed from the core
  • ABTR fuel assembly dose rate is dominated by the fuels gamma sources
  • Criticality analyses of electrorefiner show keff << 0.95 in all considered configurations
  • Shielding and cooling may be required for the liquid waste salt from electrorefiner 56

Summary & Outlook

  • Additional information is needed for improved analysis:
  • Detailed information on the salt mixtures during reprocessing
  • Onsite storage of fresh and irradiated fuel assemblies (storage containers, storage configuration, etc.)
  • Commercial size transportation canisters for UF6, reprocessed PWR and SFR fuel, spent SFR fuel
  • Related future development in SCALE:
  • Development of a strategy for SFR equilibrium core generation
  • Efficient reactivity feedback calculation
  • Integration of simple thermal expansion model
  • Reference SCALE ABTR 3D models are available online
  • Vol. 5 models will be added soon 57

SAND2023-08740PE Demonstration of MELCOR for SFR Fuel Cycle Analysis KC Wagner, David L. Luxat

MELCOR Application to Fuel Cycle Safety Assessment MELCOR is used in the DOE complex for facility safety analysis MELCOR has general and validated models for thermal hydraulic behavior of enclosures and hazardous material transport

  • Enables modeling of potential for fission products to be released from an enclosure to the environment MELCOR has been applied to safety basis development for a broad range of facility accidents that can lead to accident release of hazardous material
  • Inadvertent nuclear criticality events
  • Explosions
  • Broad range of facility fires
  • Radioactive material spills and drops MELCOR enables assessment of a range of conditions that can impact hazardous material release to the environment
  • External winds promoting enhanced transport from an enclosure to environment
  • Retention of hazardous material in filters
  • Removal of hazardous material from enclosure atmospheres by decontamination sprays Recent NRC research application of MELCOR to demonstration of safety assessment at Barnwell reprocessing facility 59

Modeling SFR Accidents with MELCOR

  • SFR materials
  • U-10Zr metallic fuel, HT-9 cladding, and sodium bond
  • SFR Fuel Representation 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

  • Reactivity accidents Reactor point kinetics and application to fast reactors
  • SFR Fuel Degradation 0.2
  • Clad pressure boundary failure, melting and candling 0.0
  • Fuel melting -0.2 Axial+radial expansion
  • Degraded fuel region molten and particulate debris behavior -0.4 U-Zr density Feedback ($)

U-Zr Doppler

  • Radionuclide release and transport -0.6 Na void Na density Gap and plenum release -0.8 CRs in CRs out Molten fuel fission gas release -1.0 Total Thermal release models -1.2
  • Sodium pool and spray fire models -1.4 1 10 100 1000 10000 100000 Time (sec)

Figure Ref. Y.I. Chang, P.J. Finck, and C. Grandy, Advanced Burner Test Reactor Preconceptual Design Report, ANL-ABR-1 (ANL-AFCI-173), 2006.] 60

Capability: Fission Product Release from SFR Fuel 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 61

Capability: Fission Product Release from Sodium Coolant Goal: Determine magnitude of fission product release into enclosure atmospheres and available to release to environment Fission product release into sodium coolant from fuel upon cladding failure

  • What fraction of fission products in the sodium are available to be released from sodium?
  • Chemical interaction of fission products with sodium critical to determine volatility of fission products Distribution of fission products in sodium influences transport out of sodium
  • Gaseous in sodium Haga et. al., Nuclear Technology 97, 177 (1992)
  • Deposited on structures interfacing with sodium Transport paths out of working fluid like sodium being considered in development
  • Evaporation influenced by solubility and vapor pressure
  • Bubble transport and bursting
  • Mechanical mobilization through jet breakup and splashing 62

Capability: Sodium Fire Modeling and Impact on Fission Product Mobilization and Transport Sodium reacts with oxygen and water 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
  • Benchmarks to ABCOVE AB5 and AB1 tests

[Figure adapted from ANL-ART-3]

63

Containment and Reactor Building ABTR defense in depth features included in the MELCOR modeling -

  • Primary containment boundary Reactor vessel Reactor vessel enclosure (top closure of the vessel with refueling port)

Intermediate heat exchanger tubes Direct Reactor Auxiliary Cooling System (DRACS) heat exchanger tubes Sodium purification piping and components

  • Secondary reactor building boundary Reactor guard vessel (nitrogen-inerted)

Reactor containment dome Sodium-to-CO2 heat exchangers DRACS intermediate system piping and systems Stainless steel-lined compartments around the vessel Purification system cell confinement Reactor building 64

Containment and Reactor Building cv Containment dome Leakage to Key sodium support systems environment

  • Argon cover gas purification Stack system cv-46 ABTR design leak rate is consistent with LWR Rails containments Leakage to
  • 0.1% vol/day at 10 psig Reactor Building HVAC environment Fan HEPA Pre-filter (design pressure)
  • Dome = 5,580 m3 cv-45 cv-44 cv-43 cv-42 Exhaust Supply Intra-building Transfer Tunnel HEPA-filtered ventilation system cv Other Reactor Building Rooms Printed
  • 2X air exchanges per hour Circuit Heat Exchanger (assumed) cv Sodium Purification Room
  • Maintains -2 H2O reactor Intermediate loop dump tank building pressure cv Air Gap 110 - Cold Pool #1 cv Guard Vessel cv Reactor Cavity 65

Scenarios Fuel Unloading Machine (FUM) failure scenario

  • Cask drop with leak in the containment dome Sodium purification pipe break during operations with coincident fuel clad failure and activity release
  • Use integrated primary system core damage models with equivalent of 217 fuel rod clad failures (i.e., 1 assembly)
  • Sodium fire in the Sodium Purification room Argon cover gas piping failure with coincident fuel clad failure and activity release
  • Use integrated primary system core damage models with equivalent of 1-assembly clad failures
  • Contaminated argon discharges into the Sodium Purification room Reprocessing accident scenario capability discussion
  • Illustrations from Barnwell safety analysis for pyro-refining or fuel fabrication plants 66

Fuel Unloading Machine (FUM) failure scenario FUM is used to load, unload, and move fuel

  • The FUM connects to the reactor enclosure for refueling operations
  • The ABTR in-vessel fuel rack can hold 36 assemblies
  • Recently discharged fuel is moved into racks for in-vessel storage (IVS)
  • Fuel remains in IVS for ~7 fuel cycles (~28 months)
  • FUM moves used fuel storage vault via the intra-building transfer tunnel MELCOR fuel damage model used to represent in the FUM SCALE provided fuel radionuclide inventories
  • HALEU spent fuel after in-vessel storage (IVS)
  • Inner Transuranic (TRU) fuel after IVS
  • Outer TRU fuel after IVS
  • HALEU fuel after irradiation 67

Fuel Unloading Machine (FUM) failure scenario Accident scenario assumptions

  • High and low leaks in FUM cask
  • Reactor building HVAC is filtering the containment dome during refueling operations
  • No residual sodium in the cask
  • All active cooling systems have failed
  • Last case uses a fuel assembly accidentally removed with only 1-day cooling after last irradiation 68

FUM accident scenario

  • During removal from the reactor, the fuel assemblies are blown dry with argon gas o No residual sodium was included in the accident scenario
  • Fuel assemblies with normal in-vessel storage cool in the damaged FUM (i.e., very low decay heat)
  • The accidental removal of a recently discharged assembly would lead to fuel failure after 40 min Fuel Temperature in the FUM Assembly Decay Heat in the FUM 1900 100000 Haleu - 1 day after irradiation Haleu - 1 day after irradiation 1700 TRU - Inner FA after IVS TRU - Inner FA after IVS TRU- Outer FA after IVS TRU - Outer FA after IVS Haleu - After IVS Haleu - After IVS 1500 Assembly decay heat (W) 10000 1300 Temperature (K) 1100 FA = Fuel assembly FUM = Fuel unloading machine IVS = In-vessel storage 900 TRU = Transuranic fuel 1000 700 FA = Fuel assembly FUM = Fuel unloading machine 500 IVS = In-vessel storage 300 100 0.1 1 10 100 1000 10000 100000 1 10 100 1000 10000 100000 Time (s) Time (s) 69

FUM accident scenario - recently discharged assembly results

  • If a recently discharged assembly is 2200 accidentally removed, it will rapidly heat 2000 Clad melting Start of fuel Fuel melting candling and to cladding candling and fuel rod failure Level 10 collapse conditions 1800 Level 9 Level 8 Level 7
  • The assembly successively relocates 1600 Level 6 Temperature (K) downward to the bottom of the storage Level 5 Level 4 cask 1400 Level 3 Level 2 Level 1
  • The high temperature fuel debris could 1200 Debris reflector fail the cask and spill out Debris Inlet FUM bottom
  • Cask failure requires further design 1000 details 800
  • Fission product release from the cask occurs through the assumed cracks after 600 being dropped 0.1 1 10 100 Time (s) 1000 10000 100000 Debris relocation to the bottom of the fuel cask 70

FUM accident scenario 0.9 Xe

  • Noble gases were rapidly released from the FUM 0.8 Containment airborne + settled Cs Ba following the fuel degradation and vented to the Iodine 0.7 Te Ru environment Mo Fraction of FA Inventory (-)

0.6 Ce La

  • Early release of more volatile cesium was captured on 0.5 U

Cd the filters 0.4 Ag

  • CsI (and NaI) and Te primarily came out following the 0.3 failure of the assembly inlet structure at 38,000 sec (10 hr) 0.2 0.1
  • HEPA filter performance modeled to degrade below 0.3 µm diameter aerosols per typical HEPA specifications 0 0 20000 40000 60000 80000 100000 Time (sec) 0.9 1.0 Xe 0.8 Xe Cs Captured on filters Cs Environment Ba Iodine Ba Te 0.7 Iodine Ru Mo Fraction of FA Inventory (-) Fraction of FA inventory (-)

Te 0.6 Ru Ce Mo Debris relocation to La U

0.5 Ce the bottom of the Cd Ag La U

0.1 fuel cask 0.4 Cd Ag 0.3 0.2 0.1 Aerosol mass median dia 0.15 to 0.4 µm at HEPA inlet 0.0 0.0 0 20000 40000 60000 80000 100000 0 20000 40000 60000 80000 100000 Time (sec) Time (sec) 71

FUM accident scenario sensitivity calculations

  • The earliest timing of an assembly removal
  • Increasing the bottom leakage flow area had a from the vessel was uncertain negligible impact on the accident scenario progression
  • Fuel collapse started at 2360 sec (0.7 hr) with Convective cooling due to leakage had a one day of cooling but increased to 8560 sec negligible impact (2.4 hr) with 10 days of cooling The upper leakage path from the FUM was much larger than the assembly flow area The base bottom leakage was equal to the assembly flow area.

Fuel temperature as a function of Fuel temperature as a function Fueltime after irradiation Temperature in the FUM Fuel Temperature of leakageinarea the FUM 1900 1900 1700 1700 Haleu - 0.1X area Haleu - Base Haleu - 10X area 1500 1500 Haleu - 100X area Haleu - 1 day after irradiation 1300 Temperature (K) 1300 Temperature (K)

Haleu - 2 days after irradiation Haleu - 3 days after irradiation 1100 Haleu - 4 days after irradiation 1100 FA = Fuel assembly Haleu - 5 days after irradiation FUM = Fuel unloading machine Haleu - 6 days after irradiation IVS = In-vessel storage 900 900 Haleu - 7 days after irradiation Haleu - 8 days after irradiation Haleu - 9 days after irradiation 700 700 FA = Fuel assembly Haleu - 10 days after irradiation Varying FUM bottom leak area -

FUM = Fuel unloading machine Base area approximately equal to assembly flow area IVS = In-vessel storage 500 500 300 300 0 2000 4000 6000 8000 10000 0 2000 4000 6000 8000 10000 Time (s) 72 Time (s)

Sodium purification system pipe break scenario The ABTR sodium purification system filters sodium from the reactor to remove hydrogen and oxygen impurities and monitors for crystallization and plugging indicators

  • The inlet and exit piping penetrates through the reactor vessel enclosure (i.e., the vessel upper lid)
  • The purification piping was specified as a 3 diameter pipe and assumed to break in the sodium purification room
  • MELCOR predicted the sodium siphon flow to be 18 kg/s with vessel cover gas pressure of 0.3 bar and a full pipe break Leakage to environment The scenario includes failure of the cladding boundary on 217 fuel rods (i.e., 1 assembly) Exhaust Supply Intra-building Transfer Tunnel The reactor building HVAC system is operating with

~2X air-changes per hour to maintain a -2 H2O gauge pressure in the sodium purification room cv Sodium Purification Room Air Gap 110 - Cold Pool #1 cv Guard Vessel r Cavity 73

cv Containment dome Sodium purification system pipe break Leakage to environment scenario Stack cv-46 Accident scenario assumptions

  • Sodium piping is isolated at 60 sec (nominally) Rails
  • Pipe break is 1 m above the floor Leakage to Reactor Building HVAC environment Fan HEPA Pre-filter
  • Siphon flow for full pipe break is 18 kg/s (varied) cv-45 cv-44 cv-43 cv-42 Exhaust Supply Intra-building Transfer Tunnel
  • Spray droplet size varied
  • Pool and spray+pool fire cv Other Reactor Building Rooms scenarios Printed Circuit Heat Exchanger cv Sodium Purification Room Intermediate loop dump tank cv Air Gap 110 - Cold Pool #1 cv Guard Vessel cv Reactor Cavity 74

Sodium purification system pipe break scenario

  • 1080 kg of sodium spilled into the purification room before being isolated o Purification system isolated at 60 sec
  • Pool fire scenario results below assume no spray oxidation and a maximum pool diameter of 3 m (i.e., room constraints)
  • Oxide layer forms on the pool surface and limits oxygen diffusion into the pool (~10% burned in 2.8 hr)
  • Pool will slowly burn for days without mitigation 1200 850 Purification room 1000 750 HEPA filter inlet Fan inlet Sodium pool Mass burned 800 650 Mass spilled Na burned (kg) Temperature (K) 600 Pool fire results 550 Pool fire results 400 450 200 350 0 250 0 2000 4000 6000 8000 10000 1 10 100 1000 10000 Time (s) Time (s) 75

Sodium purification system pipe break scenario 2500

  • The sodium burn rate is controlled by the oxide layer on the 0.1x diffusivity pool surface 2000 Base diffusivity 10x diffusivity o Oxide layer eventually builds up to limit the burn rate 100x diffusivity Temperature (K) 1500
  • Oxygen diffusivity across the oxide layer on the pool surface has uncertainties, which initially affect the burn rate 1000 o e.g., pool geometry, pool temperature, room oxygen o Oxide layer eventually limits oxygen diffusivity 500 Sodium purification
  • The peak room temperature and the gas temperature to the room temperature HEPA filters is strongly impacted by the initial burn rate 0 1 10 100 1000 10000 Time (s) 250 1200 0.1x diffusivity 0.1x diffusivity Base diffusivity Base diffusivity 10x diffusivity 1000 200 10x diffusivity 100x diffusivity 100x diffusivity Pool fire 800 results Na burned (kg) Temperature (K) 150 Mass of sodium 600 Inlet temperature to burned the HEPA filters 100 400 50 200 0 0 0 2000 4000 6000 8000 10000 1 10 100 1000 10000 Time (s) Time (s) 76

Sodium purification system pipe break scenario

  • Sodium fires generate lots of aerosols o 2 Na + 1/2 O2 Na2O (dominant in these calculations)
  • Sodium byproduct aerosols plug filters and reduce HVAC flow & effectiveness o Base case assumes 1 HEPA filter unit (i.e., not described in the ABTR reference report) o Sensitivity calculations assess the impact of 2, 4, and 8 HEPA filter units 80 4.0 Na2O generated 70 Base HEPA Pool fire 3.5 HVAC flowrate Base HEPA 2X HEPA 2X HEPA results 4X HEPA 4X HEPA 60 8X HEPA 3.0 8X HEPA Total filter mass (kg) 50 Na2O generated HVAC Flow (m3/s) 2.5 and filtered mass 40 2.0 30 1.5 20 1.0 10 0.5 0 0.0 0 2000 4000 6000 8000 10000 0 2000 4000 6000 8000 10000 Time (s) Time (s) 77

Sodium purification system pipe break scenario

  • Next examples include combined spray and pool fires o Includes spray interaction with the room oxygen with continuation in a pool fire
  • Base case is 18 kg/s with a large droplet size (i.e., characteristic of low-pressure pour)
  • Other cases explored smaller spray droplet sizes, smaller flowrates, and isolated or not isolated o Mass burned is a function of droplet size, leak rate, and leak duration Spray base case 20000 160 Spray 0.5X droplet size Spray base case Spray 0.5X droplet size, 0.5X flowrate 18000 Spray 0.5X droplet size 140 Spray 0.01X droplet size, 0.01X flowrate Spray 0.5X droplet size, 0.5X flowrate Spray 0.001X droplet size, 0.001X flowrate 16000 Spray 0.01X droplet size, 0.01X flowrate Spray 0.001X droplet size, 0.001X flowrate 120 14000 Combined spray 100 Combined spray Na burned (kg) 12000 Na spilled (kg) and pool fire and pool fire 10000 80 2 cases not isolated 8000 60 Mass burned 6000 40 4000 Mass spilled 20 2000 0 0 0 2000 4000 6000 8000 10000 0 2000 4000 6000 8000 10000 Time (s) Time (s) 78

Sodium purification system pipe break scenario

  • Spray fire room temperatures can be much higher due to the spray burn efficiency versus a pool fire (i.e., function of droplet size, fall height, spray velocity)
  • Sodium fires can be oxygen limited (HVAC remains operational) o Contrast the 0.001X spray droplet results at 0.001X mass flow rate with base case response 2500 0.25 Spray base case Spray base case Spray 0.5X droplet size Spray 0.5X droplet size Spray 0.5X droplet size, 0.5X flowrate Spray 0.5X droplet size, 0.5X flowrate 2000 0.2 Spray 0.01X droplet size, 0.01X flowrate Spray 0.01X droplet size, 0.01X flowrate Spray 0.001X droplet size, 0.001X flowrate Spray 0.001X droplet size, 0.001X flowrate Oxygen concnetration (-)

Combined spray Combined spray Temperature (K) 1500 0.15 and pool fire and pool fire 1000 0.1 Room temperature Oxygen concentration 500 0.05 0 0 0 2000 4000 6000 8000 10000 0 2000 4000 6000 8000 10000 Time (s) Time (s) 79

Sodium purification system pipe break scenario

  • Release magnitude is limited by (a) the small amount of radionuclide inventory in the spill and (b) the slow burning rate (i.e., release rate is proportional to burn rate)
  • The airborne concentration steadily decreases due to HVAC flow (initially 2 room changes per hour)
  • HEPA filter captures most radionuclides and limits environmental release 1.E-06 1.E-05 1.E-07 1.E-06 Airborne in the sodium purification room 1.E-07 1.E-08 Xe Environmental release 1.E-08 Cesium Fraction of inventory (-) Fraction of inventory (-)

1.E-09 Ba Iodine 1.E-09 Te 1.E-10 Ru Xe Cesium 1.E-10 Mo 1.E-11 Ba Ce Iodine 1.E-11 La Te U 1.E-12 Cd Ru 1.E-12 Mo Ag 1.E-13 Ce La 1.E-13 U

1.E-14 Cd 1.E-14 Ag 1.E-15 1.E-15 0 2000 4000 6000 8000 10000 0 2000 4000 6000 8000 10000 Time (sec) Time (sec) 80

cv Containment dome Cover-gas pipe break scenario Leakage to environment Stack Accident scenario assumptions cv-46

  • Cover-gas piping is not isolated
  • Discharge flow is steady and Rails maintained by large pressure Leakage to control supply tanks Reactor Building HVAC environment Fan HEPA Pre-filter
  • HVAC is running with 2X air changes per hour Exhaust cv-45 cv-44 cv-43 cv-42 Supply Intra-building Transfer Tunnel
  • The scenario includes failure of the cladding boundary on 217 fuel cv Other Reactor Building Rooms rods (i.e., 1 assembly) Printed Circuit Heat cv Sodium Purification Room Exchanger Intermediate loop dump tank cv Air Gap 110 - Cold Pool #1 cv Guard Vessel cv Reactor Cavity 81

Cover-gas pipe break scenario

of the sodium pool

  • Most of the NaI remains in the pool due to its low vapor
  • Once in the cover gas, they leak through the cover gas pressure in this scenario (~0.01 Pa) pipe break.
  • The released NaI condenses into small aerosols that
  • The HVAC circulates the released gases out the plant stack are not completely filtered by the HEPA 1.0E+00 1.0E+00 1.0E-01 Released Noble gas behavior 1.0E-01 Released NaI behavior 1.0E-02 1.0E-02 Released 1.0E-03 Released 1.0E-03 In-vessel 1.0E-04 In-vessel Fraction of inventory (-)

Reactor building Fraction of inventory (-)

Reactor building 1.0E-04 1.0E-05 Environment Filters 1.0E-06 Environment 1.0E-05 1.0E-07 1.0E-06 1.0E-08 1.0E-07 1.0E-09 1.0E-10 1.0E-08 1.0E-11 1.0E-09 1.0E-12 1.0E-10 1.0E-13 1 10 100 1000 10000 1 10 100 1000 10000 Time (sec) Time (sec) 82

Examples for fuel fabrication and reprocessing safety analysis

Reprocessing and fuel fabrication accident analysis BNFP facility drawing - Supply-side ventilation Safety-grade ventilation 7

Hot cells for hazardous material processing and filtration system

-0.49 kPa

-0.12 kPa 0 kPa

-0.12 kPa 6 BNFP facility drawing The processing (hot) cells are where the fire and/or explosion events are simulated. The regions are enclosed with red dotted lines There is generally a flow from the least radioactive regions Hot cells were the fire towards the hot cells, which are the contain the processes with and/or explosion events the highest radioactive inventories.

are simulated 8 BNFP facility drawing - Exhaust-side ventilation The plant stack is the filtered release pathway after 2 sets of filters The processing (hot) cells are where the fire and/or explosion events are simulated. The regions are enclosed with red dotted lines Ref. [K. C. Wagner and David L.Y. Louie, MELCOR Demonstration Analysis Of Accident Scenarios At A Spent Nuclear Reprocessing Plant, 28th International Conference on Nuclear Engineering, August 2- 6, 2020, Anaheim, CA, USA, ICONE28-POWER2020-16584] 84

Reprocessing and fuel fabrication accident analysis 12 Examples of Fire Scenario Results - Radionuclide Results Sensitivity of the Fire Size Modeling Accident - Key boundary conditions Fans draw released radionuclides to the filters, Example of an explosion scenario where they are captured by the HEPA filters Activity Distribution 1.E+13 14 Example of an Explosion Scenario Result 1.E+12 Activity distribution in the first hour after

  • Pressure response figure below the accident Activity Release (Bq) 1.E+11 shows immediate failure of HEPA Activity Distribution The environmental release is Environment Filter 7 at the exit of the PPC 120%

Environment relatively small because the HEPA 1.E+10 filters remained intact. The activity Exhaust Total Hot Cells

  • The dissipation of the pressure from 100%

Total Exhaust Filter 1 release is due to aerosols below the min. effective HEPA capture size Support Gallery the explosion also fails the final Hot Cells Support Gallery Activity Distribution (%)

exhaust filter within 13 seconds 80%

and radionuclide gases. Other Other 1.E+09 60%

Pressure response at the filters between the PPC and the stack 30 40%

1.E+08 HEPA Filter 4 0 3 6 9 12 HEPA Filter 1 20%

Time (hr) 25 AFS & VFS always <0 psig AFS VFS End of the explosion 0%

20 0 600 1200 1800 2400 3000 3600 Pressure drop (kPa)

Time (sec)

  • Activity distribution above shows a HEPA Filter 4 fails 15 HEPA Filter 1 fails large release to the environment due Example of a fire scenario to the failure of the two HEPA 2.49 kPa = HEPA overpreessure failure 10 filters between the PPC and the 5 plant stack.
  • Pre-filter 1 remains intact and retains 0

-5 0 5 10 15 show larger aerosols Time (sec)

Ref. [K. C. Wagner and David L.Y. Louie, MELCOR Demonstration Analysis Of Accident Scenarios At A Spent Nuclear Reprocessing Plant, 28th International Conference on Nuclear Engineering, August 2- 6, 2020, Anaheim, CA, USA, ICONE28-POWER2020-16584] 85

Reprocessing and fuel fabrication accident analysis Argonne National Laboratory and Merrick & Company, Engineering Services recently published a concept for a pyro-processing plant

  • Insufficient information for a demonstration calculation
  • Similar to the Barnwell facility, work done in hot cells
  • Cited limiting accident with oxidation of 1000-2000 kg of uranium metals
  • Other accidents due to loss of heat removal for TRU vault
  • Fuel fabrication could include spill accidents during casting and alloying steps

[Yoon Il Chang, et al. (2018): Conceptual Design of a Pilot-Scale Pyroprocessing Facility, Nuclear Technology https://doi.org/10.1080/00295450.2018.1513243]

86

MELCOR Summary MELCOR SFR Summary

  • MELCOR capabilities were demonstrated New phenomenological modeling added to MELCOR for SFRs Application of radionuclide transport models
  • Capabilities for a range of SFR fuel cycle accident scenarios
  • Key physics considered SFR assembly thermal hydraulics Sodium fires Fission product release
  • Future work Fission product release modeling from spills and sodium fires Radionuclide chemistry 88

Workshop Summary Closing Remarks

  • Demonstration of NRCs Code Readiness for Simulating non-LWRs

- HTGR Nuclear Fuel Cycle (Completed February 2023)

- SFR Nuclear Fuel Cycle (Today)

  • Next Steps

- Public Reports

  • Coming in 2023, Non-LWR Fuel Cycle Scenarios for SCALE and MELCOR Modeling Capability Demonstration

- MSR Nuclear Fuel Cycle Workshop (2024) 90

Backup IAEA -TECDOC-2006 Notes CDA no mixing release insights

  • No mixture compound vapor pressure Halogens
  • Ideal mixture Raoults Law o NaI(l) is predominant chemical species for real mixture
  • Real mixture Excess for deviation from o Bromine forms CsBr with 50% release at 950 K (no mixture) Raoults Law but drops to <10-4 with mixing Alkali metals o Cs binds to CsI, CsRb. CsBr, CsNa with 90% release No mixture assumption o Complete Rb release Tellurium o BaTe which does not release Others o Noble metals are solid & do not release o Lanthanides form oxides and dependent on oxygen availability o Eu is volatile (13% release) if it does not form Eu2O3 o Ce, Pu, and Np are stable 92

IAEA -TECDOC-2006 Notes

  • No mixture compound vapor pressure
  • Ideal mixture Raoults Law
  • Real mixture Excess for deviation from Raoults Law 873 K No mixture assumption 93
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