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
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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 Material Safety and Safeguards Office of Nuclear Reactor Regulations 1

  • 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 Outline 2

NRCs Strategy for Preparing for non-LWRs NRCs Readiness Strategy for Non-LWRs Phase 1 - Vision & Strategy Phase 2 - Implementation Action Plans IAPs are planning tools that describe:

Required work, resources, and sequencing of work to achieve readiness Strategy #2 - Computer Codes and Review Tools Identifies computer code & development activities Identifies key phenomena Assess available experimental data & needs IAP Strategy #2 Computer Codes and Tools Volume #1 Systems Analysis Volume #2 Fuel Performance Volume #3 Source Term, Consequence Volume #4 Licensing &

Dose Volume #5 Nuclear Fuel Cycle 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 Uranium Recovery / Milling - 10 CFR Part 20 Uranium Conversion - 10 CFR Parts 30, 40, 70, 73 and 76 Uranium Enrichment - 10 CFR Parts 30, 40, 70, 73 and 76 Fuel Fabrication - 10 CFR Parts 30, 40, 70, 73 and 76 Reactor Utilization - 10 CFR Parts 50 & 74 Spent Fuel Pool Storage - 10 CFR Parts 50.68 Spent Fuel Storage (Dry) - 10 CFR Parts 63, 71, and 72 5

Project Scope - Non-LWR Fuel Cycle Enrichment UF6 enrichment UF6 Transportation Fuel Fabrication Fresh Fuel Transportation Fuel Utilization (including on-site spent fuel storage)

  • Not envisioned to change from current methods.

Uranium Mining & Milling

  • Successfully completed and leveraged from the Volume 3 - Source Term & Consequence work Power Production
  • Large amount of uncertainties for non-LWR concepts & lack of information Spent Fuel Off-site Storage & Transportation
  • Large amount of uncertainties for non-LWR concepts & lack of information Spent Fuel Final Disposal Stages in scope for Volume 5 Stages out of scope for Volume 5 6

Codes Supporting non-LWR Nuclear Fuel Cycle Licensing

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

Project Approach Build representative fuel cycle designs leveraging the Volume 3 designs Identify key scenarios and accidents exercising key phenomena & models Build representative SCALE & MELCOR models and evaluate Code Assessment Representative Initial and Boundary Conditions Simulating Accidents around Key Phenomena Sensitivity Studies Identify &

Address Modeling Gaps 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

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 of the SFR fuel cycle F. Bostelmann

11 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 Overview

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

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

14 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 E1: Enrichment Major hazards:

UF6 liquid and vapor leaks from damaged pipes or cylinders Criticality due to unintended accumulation of enriched U

15 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:

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 T1: Transportation of UF6 Ref.: ORANO Safety Analysis Report for the DN30-X Package https://www.nrc.gov/docs/ML2232/ML22327A183.pdf Certificate of Compliance, Certificate number 9388 https://rampac.energy.gov/docs/default-source/certificates/1019388.pdf Package design Enrichment limit Permissible UF6 mass DN30-10 10 wt.% 235U 1460 kg DN30-20 20 wt.% 235U 1271 kg Major hazards:

Criticality due to water accidents and container drop Release of UF6 due to container rupture DN30-X package 30B-X cylinder

16 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 R1: Reprocessing of Spent Nuclear Fuel Refs.:

[1] 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.

[2] 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.

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

Schematic of the electrometallurgical treatment used for metallic fuel from the EBR-II Major hazards:

Criticality from misfeeding or mishandling of fuel Release of radiological materials

17 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 F1: Fabrication of Metallic Fuel 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.

18

1. Fuel rod fabrication:

Fuel cladding tube is fabricated and cleaned Cladding tube is loaded with sodium to facilitate bonding Fuel slugs are loaded into the cladding tube Fuel cladding tube is closure welded to achieve sealing

2. Fuel assembly manufacturing F2: Fabrication of Fuel Assemblies Major hazards:

Release of hazardous or corrosive chemicals/gases Criticality from misfeeding or mishandling of fuel Release of radiological materials or sodium from rods Ref.: D. E. Burkes, et al. A US Perspective on Fast Reactor Fuel Fabrication Technology and Experience Part 1: Metal Fuels and Assembly Design. Journal of Nuclear Materials, 389:458-469, 2009.

19 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 T2: Transportation of Fresh Fuel Assemblies to Plant ES-3100 Ref.: J. Jarrell, A Proposed Path Forward for Transportation of High-Assay Low-Enriched Uranium, INL Technical Report, INL/EXT-18-51518 Rev 0 (2018).

Major hazards:

Criticality due to water accidents and container drop Corrosion of sodium bond Reaction of sodium with water, air, or concrete in case of container ruptures

20 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 U1/U2/U4 - Utilization Stages Ref.: Y. I. Chang, et al., Advanced Burner Test Reactor Preconceptual Design Report, ANL-ABR-1/ANL-AFCI-173, Argonne National Laboratory, 2006.

21 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 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 Intra-building casks (IBC): lead-shielded inter-building casks with inert gas atmosphere, with or without active cooling Intra-building transfer tunnel: transfer of assemblies within inter-building cask U1/U2/U4: Major Components for Fuel Handling Pantograph Rotatable plug Storage rack Ref.: Y. I. Chang, et al., Advanced Burner Test Reactor Preconceptual Design Report, ANL-ABR-1/ANL-AFCI-173, Argonne National Laboratory, 2006.

22 U1/U2/U4: Major Hazards Major hazards:

Reaction of sodium with water, air, or concrete Corrosion of sodium bond 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.

23 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 Use of sodium bond and sodium coolant 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 Summary

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 Demonstration of SCALE for SFR Fuel Cycle Analysis D. Hartanto

25 OBJECTIVE AND APPLICATIONS Ref.: Chang, Y. I., et al. Advanced Burner Test Reactor -

Preconceptual Design Report. Technical Report ANL-ABR-1 (ANL-AFCI-173), Argonne National Laboratory, 2006.

ABTR reactor building

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

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

  • Accident: A leak in the waste stream storage tank allows for release of fission products during reprocessing.
  • Analysis: Determine fuel inventory and perform SCALE activity calculations.

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

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

26 OBJECTIVE AND APPLICATIONS Ref.: Pyroprocessing Technologies Brochure, Argonne National Laboratory Electrorefiner Objective: Demonstrate use of SCALE for simulating accident scenarios in all stages of the nuclear fuel cycle for Sodium-cooled Fast Reactors (SFR)

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

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

  • Accident: A leak in the waste stream storage tank allows for release of fission products during reprocessing.
  • Analysis: Determine fuel inventory and perform SCALE activity calculations.

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

27 Advanced Burner Test Reactor (ABTR)

REFERENCE SODIUM FAST REACTOR DESIGN Reactor Power 250 MWt, 95 MWe Coolant Temperature 355°C/510°C Fuel Metallic Cladding and Duct HT-9 Cycle Length 4 months 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.

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

Inner FA Mid FA Outer FA SCALE ABTR Model

28 APPLIED SCALE6.3.1 SEQUENCES Rapid inventory generation with ORIGAMI

  • Depletion and decay solver (ORIGEN)
  • Requires pre-calculated ORIGEN cross-section libraries (generated in previous work for the ABTR*)
  • Output:

Nuclide inventory of irradiated fuel 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)

Shielding & radiation dose calculations with MAVRIC

  • Monte Carlo photon and neutron transport code (MONACO) with automated variance reduction for shielding analyses
  • Requires radiation source terms.
  • Output:

- Spatial flux/dose rate distributions Criticality calculation with CSAS

  • Monte Carlo neutron transport code (KENO or Shift) for criticality safety analysis
  • Output:

- Multiplication factor

- Spatial flux and fission density distributions Nuclide inventory and decay heat of the irradiated fuel are passed to MELCOR.

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, 2023.

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

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

U/TRU-10Zr 16.5 wt.% (inner) & 20.7 wt.% TRU (outer)

Specific power: 65.6 GW/tHM (inner) & 51.4 GW/tHM (outer)

Discharged BU: 94.5 GWd/tHM (inner) & 92.6 GWd/tHM (outer)

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

Gas plenum 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.

ABTR fuel assembly

30 SPENT NUCLEAR FUEL Rapid inventory generation with ORIGAMI Irradiation history:

TRU Inner Loaded for 12 cycles 120 days per cycle TRU Outer Loaded for 15 cycles 120 days per cycle HALEU Loaded for 6 cycles 540 days per cycle Assuming 10 days of cooling time between cycles Discharged fuel assembly is planned to be stored for 7 reactor cycles in the in-vessel storage (IVS)

Fuel Na bond Lower refl.

Gas plenum ABTR fuel assembly

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

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

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

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

106Rh (12%)

144Pr (9%)

95Nb (8%)

95Zr (8%)

137mBa (22%)

106Rh (14%)

90Y (12%)

238Pu (9%)

134Cs (7%)

U/TRU Outer 140La (21%)

106Rh (12%)

144Pr (9%)

95Nb (8%)

95Zr (8%)

137mBa (22%)

106Rh (12%)

90Y (12%)

238Pu (10%)

134Cs (6%)

HALEU 140La (21%)

144Pr (11%)

95Nb (9%)

95Zr (9%)

106Rh (7%)

90Y (29%)

137mBa (29%)

134Cs (11%)

137Cs (7%)

238Pu (6%)

PWR*

140La (21%)

144Pr (10%)

95Nb (8%)

106Rh (8%)

95Zr (8%)

90Y (25%)

137mBa (25%)

238Pu (11%)

244Cm (11%)

137Cs (7%)

  • Decay heat at shutdown is similar between the different fuel types (~5-7% power)
  • Initially, slightly higher for the U/TRU inner fuel due to higher specific power although its burnup is lower than HALEU

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

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

103mRh (8%)

95Nb (7%)

95Zr (7%)

141Ce (6%)

137Cs (19%)

137mBa (18%)

241Pu (15%)

147Pm (12%)

90Y (7%)

U/TRU Outer 103Ru (8%)

103mRh (8%)

95Nb (7%)

95Zr (6%)

141Ce (6%)

137Cs (19%)

241Pu (19%)

137mBa (18%)

147Pm (11%)

90Y (7%)

HALEU 95Nb (8%)

95Zr (7%)

103Ru (6%)

103mRh (6%)

141Ce (6%)

137Cs (22%)

137mBa (21%)

90Y (16%)

90Sr (16%)

147Pm (10%)

PWR*

140La (32%)

95Nb (14%)

95Zr (13%)

103Ru (9%)

134Cs (6%)

137mBa (76%)

134Cs (16%)

154Eu (7%)

125Sb (0.5%)

106Rh (0.2%)

  • Similar trends compared to decay heat
  • PWR has the lowest activity due to lower FPs built-up

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

35 CONTAINMENT BUILDING (CB) MODEL MAVRIC model of the CB and unshielded fuel assembly (front view) 1.2-cm thick steel liner Reinforced concrete

(~1 m) assuming rebar-to-concrete mass ratio of 0.106 Fuel assembly 3D view of the CB with front quarter segment removed Fuel assembly 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 Radiation dose rate inside and outside of containment are calculated with MAVRIC using intact fuel assembly as radiation source (irradiated fuel and activation products).

ANSI standard (1977) flux-to-dose-rate factors Cartesian and cylindrical mesh for dose calculations Statistical error < 0.5%

Refs.:

[1] P. F. Peterson et al., Metal and Concrete Inputs for Several Nuclear Power Plants, Report UCBTH-05-001, 2005.

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

36

  • Neutron sources from spontaneous fission

Fuel light element impurities might contribute additional neutron sources Cooling time (d)

HALEU TRU NEUTRON SOURCE TERMS Cooling time ABTR HALEU ABTR U/TRU (inner) 10 days Cm-242 (74.2%)

Pu-240 (17.3%)

Cm-242 (44.3%)

Cm-244 (54.0%)

7 cycles Pu-240 (71.3%)

Pu-238 (16.2%)

Cm-244 (11.5%)

Cm-244 (94.1%)

Cm-242 (0.27%)

Half life:

Cm-242: 162.8 d Cm-244: 18.10 y 7 cycles of cooling time:

U/TRU: 840 d HALEU: 3780 d

37

  • 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)

GAMMA SOURCE TERMS

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

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

Position 1 ABTR HALEU fuel assembly leaning on the containment wall (front view)

  • Highest dose rate observed when fuel assembly leans on containment wall This model is used for all dose rate calculations 1

2 Location of the FA

39 10 days of cooling MAIN BETA AND GAMMA EMITTERS Nuclide Half-life Nuclide Half-life Y-91 58.5 d Cs-137/Ba-137m 30.07 yr/2.552 m Zr-95 64.02 d Ba-140 12.75 d 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 Ru-106/Rh-106 1.02 yr/2.18 h Nd-147 10.98 d Sb-124 60.2 d Pm-148m 42.3 d Te-132/I-132 3.2 d/2.28 h Eu-154 8.593 yr Cs-134 2.065 yr Eu-156 15.2 d Cs-136/Ba-136m 13.16 d/0.308 s 7 cycles of cooling Nuclide Half-life Sr-90/Y-90 28.78 yr/2.67 d Ru-106/Rh-106 1.02 yr/2.18 h Ag-110m 249.8 d Sb-125 2.758 yr Cs-134 2.065 yr Cs-137/Ba-137m 30.07 yr/2.552 m Ce-144/Pr-144 284.6 d/17.28 m Eu-152 13.54 yr Eu-154 8.593 yr Nuclides important to the gamma source terms for both ABTR U/TRU and HALEU fuels

40 DOSE RATE MAP INSIDE CB Dose rate (rem/h) 4.6x106 rem/h (4.6x104 Sv/h) 7.0x102 rem/h (7.0 Sv/h)

ABTR HALEU 6.0x106 rem/h (6.0x104 Sv/h) 9.0x102 rem/h (9.0 Sv/h)

ABTR U/TRU 10 days cooling time

41 DOSE RATE MAP INSIDE CB Dose rate (rem/h) 9.2x104 rem/h (9.2x102 Sv/h) 13.5 rem/h (0.135 Sv/h)

ABTR HALEU 1.9x105 rem/h (1.9x103 Sv/h) 30 rem/h (0.3 Sv/h)

ABTR U/TRU 7 cycles of cooling time

42 DOSE RATE MAPS OUTSIDE CB Dose rate (mrem/h) 530 mrem/h (5.3 mSv/h) 0.4 mrem/h (4 Sv/h)

Fuel assembly ABTR HALEU ABTR U/TRU 720 mrem/h (7.2 mSv/h) 0.5 mrem/h (5 Sv/h)

Fuel assembly 10 days cooling time

43 DOSE RATE MAPS OUTSIDE CB Dose rate (mrem/h)

ABTR HALEU 0.34 mrem/h (3.4 Sv/h) 0.2 rem/h (2.0E-03 Sv/h)

Fuel assembly 6.6 mrem/h (66 Sv/h) 5 rem/h (5E-02 Sv/h)

Fuel assembly ABTR U/TRU 7 cycles of cooling time

44

  • 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)

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 (4.6x104 Sv/h) 6.0x106 rem/h (6.0x104 Sv/h) 530 mrem/h (5.3 mSv/h) 720 mrem/h (7.2 mSv/h) 7 cycles 9.2x104 rem/h (9.2x102 Sv/h) 1.9x105 rem/h (1.9x103 Sv/h) 0.34 mrem/h (3.4 Sv/h) 6.6 mrem/h (66 Sv/h)

  • 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).

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

46 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.

ELECTROMETALLURGICAL PROCESSING 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.

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

47 ELECTROREFINING 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.

Mark-IV Electrorefiner GBZ: Glass-bonded zeolite

48 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.

49 CSAS Model of Electrorefiner (40x40)

CRITICALITY ANALYSIS OF ELECTROREFINER Cadmium pool (6)

Salt (12)

LiCl-KCl-PuCl3 (FP&TRU)

Anode basket Steel cathode Pure U (Dendritic)

Single Cathode ER Dual Cathodes ER 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.

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

51 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)

CRITICALITY ANALYSIS OF ELECTROREFINER Single Cathode ER Dual Cathodes ER

52 CRITICALITY ANALYSIS OF ELECTROREFINER Single Cathode ER Dual Cathodes ER 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

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

54

  • 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.

ACTIVITY OF SALT

55 Summary

56

  • 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 Summary & Outlook

57

  • 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

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

59 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

60

  • 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
  • Clad pressure boundary failure, melting and candling
  • Fuel melting
  • Degraded fuel region molten and particulate debris behavior
  • Radionuclide release and transport Gap and plenum release Molten fuel fission gas release Thermal release models
  • Sodium pool and spray fire models Modeling SFR Accidents with MELCOR

-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 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.]

61 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

62 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
  • 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 Capability: Fission Product Release from Sodium Coolant Haga et. al., Nuclear Technology 97, 177 (1992)

63 Capability: Sodium Fire Modeling and Impact on Fission Product Mobilization and Transport

[Figure adapted from ANL-ART-3]

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

64 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

65 Containment and Reactor Building cv Other Reactor Building Rooms cv-42 cv-43 cv-44 cv-45 cv-46 Stack Fan HEPA Pre-filter Reactor Building HVAC Intermediate loop dump tank Printed Circuit Heat Exchanger Leakage to environment cv Containment dome Leakage to environment 110 - Cold Pool #1 cv Guard Vessel cv Reactor Cavity cv Air Gap Supply Exhaust Rails cv Sodium Purification Room Intra-building Transfer Tunnel Key sodium support systems

  • Argon cover gas purification system ABTR design leak rate is consistent with LWR containments
  • 0.1% vol/day at 10 psig (design pressure)
  • Dome = 5,580 m3 HEPA-filtered ventilation system
  • 2X air exchanges per hour (assumed)
  • Maintains -2 H2O reactor building pressure

66 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 Scenarios

67 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 Fuel Unloading Machine (FUM) failure scenario 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

68 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

69 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 100 1000 10000 100000 1

10 100 1000 10000 100000 Assembly decay heat (W)

Time (s)

Assembly Decay Heat in the FUM Haleu - 1 day after irradiation TRU - Inner FA after IVS TRU - Outer FA after IVS Haleu - After IVS FA = Fuel assembly FUM = Fuel unloading machine IVS = In-vessel storage TRU = Transuranic fuel 300 500 700 900 1100 1300 1500 1700 1900 0.1 1

10 100 1000 10000 100000 Temperature (K)

Time (s)

Fuel Temperature in the FUM Haleu - 1 day after irradiation TRU - Inner FA after IVS TRU-Outer FA after IVS Haleu - After IVS FA = Fuel assembly FUM = Fuel unloading machine IVS = In-vessel storage

70 FUM accident scenario - recently discharged assembly results If a recently discharged assembly is accidentally removed, it will rapidly heat to cladding candling and fuel rod failure conditions The assembly successively relocates downward to the bottom of the storage cask The high temperature fuel debris could fail the cask and spill out Cask failure requires further design details Fission product release from the cask occurs through the assumed cracks after being dropped 600 800 1000 1200 1400 1600 1800 2000 2200 0.1 1

10 100 1000 10000 100000 Temperature (K)

Time (s)

Clad melting Fuel melting Level 10 Level 9 Level 8 Level 7 Level 6 Level 5 Level 4 Level 3 Level 2 Level 1 Debris reflector Debris Inlet FUM bottom Start of fuel candling and collapse Debris relocation to the bottom of the fuel cask

71 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 0

20000 40000 60000 80000 100000 Fraction of FA Inventory (-)

Time (sec)

Xe Cs Ba Iodine Te Ru Mo Ce La U

Cd Ag 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 0

20000 40000 60000 80000 100000 Fraction of FA Inventory (-)

Time (sec)

Xe Cs Ba Iodine Te Ru Mo Ce La U

Cd Ag FUM accident scenario

  • Noble gases were rapidly released from the FUM following the fuel degradation and vented to the environment
  • Early release of more volatile cesium was captured on the filters
  • CsI (and NaI) and Te primarily came out following the failure of the assembly inlet structure at 38,000 sec (10 hr)
  • HEPA filter performance modeled to degrade below 0.3 µm diameter aerosols per typical HEPA specifications Captured on filters Containment airborne + settled Environment 0.0 0.1 1.0 0

20000 40000 60000 80000 100000 Fraction of FA inventory (-)

Time (sec)

Xe Cs Ba Iodine Te Ru Mo Ce La U

Cd Ag Aerosol mass median dia 0.15 to 0.4 µm at HEPA inlet Debris relocation to the bottom of the fuel cask

72 300 500 700 900 1100 1300 1500 1700 1900 0

2000 4000 6000 8000 10000 Temperature (K)

Time (s)

Fuel Temperature in the FUM Haleu - 0.1X area Haleu - Base Haleu - 10X area Haleu - 100X area FA = Fuel assembly FUM = Fuel unloading machine IVS = In-vessel storage Varying FUM bottom leak area -

Base area approximately equal to assembly flow area FUM accident scenario sensitivity calculations

  • The earliest timing of an assembly removal from the vessel was uncertain
  • Fuel collapse started at 2360 sec (0.7 hr) with one day of cooling but increased to 8560 sec (2.4 hr) with 10 days of cooling 300 500 700 900 1100 1300 1500 1700 1900 0

2000 4000 6000 8000 10000 Temperature (K)

Time (s)

Fuel Temperature in the FUM Haleu - 1 day after irradiation Haleu - 2 days after irradiation Haleu - 3 days after irradiation Haleu - 4 days after irradiation Haleu - 5 days after irradiation Haleu - 6 days after irradiation Haleu - 7 days after irradiation Haleu - 8 days after irradiation Haleu - 9 days after irradiation Haleu - 10 days after irradiation FA = Fuel assembly FUM = Fuel unloading machine IVS = In-vessel storage

  • Increasing the bottom leakage flow area had a negligible impact on the accident scenario progression Convective cooling due to leakage had a negligible impact 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 time after irradiation Fuel temperature as a function of leakage area

73 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 Sodium purification system pipe break scenario The scenario includes failure of the cladding boundary on 217 fuel rods (i.e., 1 assembly)

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 Leakage to environment 110 - Cold Pool #1 cv Guard Vessel r Cavity Air Gap Supply Exhaust cv Sodium Purification Room Intra-building Transfer Tunnel

74 Sodium purification system pipe break scenario cv Other Reactor Building Rooms cv-42 cv-43 cv-44 cv-45 cv-46 Stack Fan HEPA Pre-filter Reactor Building HVAC Intermediate loop dump tank Printed Circuit Heat Exchanger Leakage to environment cv Containment dome Leakage to environment 110 - Cold Pool #1 cv Guard Vessel cv Reactor Cavity cv Air Gap Supply Exhaust Rails cv Sodium Purification Room Intra-building Transfer Tunnel Accident scenario assumptions

  • Sodium piping is isolated at 60 sec (nominally)
  • Pipe break is 1 m above the floor
  • Siphon flow for full pipe break is 18 kg/s (varied)
  • Spray droplet size varied
  • Pool and spray+pool fire scenarios

75 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 0

200 400 600 800 1000 1200 0

2000 4000 6000 8000 10000 Na burned (kg)

Time (s)

Mass burned Mass spilled 250 350 450 550 650 750 850 1

10 100 1000 10000 Temperature (K)

Time (s)

Purification room HEPA filter inlet Fan inlet Sodium pool Pool fire results Pool fire results

76 0

500 1000 1500 2000 2500 1

10 100 1000 10000 Temperature (K)

Time (s) 0.1x diffusivity Base diffusivity 10x diffusivity 100x diffusivity 0

50 100 150 200 250 0

2000 4000 6000 8000 10000 Na burned (kg)

Time (s) 0.1x diffusivity Base diffusivity 10x diffusivity 100x diffusivity 0

200 400 600 800 1000 1200 1

10 100 1000 10000 Temperature (K)

Time (s) 0.1x diffusivity Base diffusivity 10x diffusivity 100x diffusivity Sodium purification system pipe break scenario The sodium burn rate is controlled by the oxide layer on the pool surface o

Oxide layer eventually builds up to limit the burn rate Oxygen diffusivity across the oxide layer on the pool surface has uncertainties, which initially affect the burn rate o

e.g., pool geometry, pool temperature, room oxygen o

Oxide layer eventually limits oxygen diffusivity The peak room temperature and the gas temperature to the HEPA filters is strongly impacted by the initial burn rate Sodium purification room temperature Inlet temperature to the HEPA filters Mass of sodium burned Pool fire results

77 0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 0

2000 4000 6000 8000 10000 HVAC Flow (m3/s)

Time (s)

Base HEPA 2X HEPA 4X HEPA 8X HEPA 0

10 20 30 40 50 60 70 80 0

2000 4000 6000 8000 10000 Total filter mass (kg)

Time (s)

Na2O generated Base HEPA 2X HEPA 4X HEPA 8X HEPA 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 Na2O generated and filtered mass HVAC flowrate Pool fire results

78 0

20 40 60 80 100 120 140 160 0

2000 4000 6000 8000 10000 Na burned (kg)

Time (s)

Spray base case Spray 0.5X droplet size Spray 0.5X droplet size, 0.5X flowrate Spray 0.01X droplet size, 0.01X flowrate Spray 0.001X droplet size, 0.001X flowrate 0

2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 0

2000 4000 6000 8000 10000 Na spilled (kg)

Time (s)

Spray base case Spray 0.5X droplet size Spray 0.5X droplet size, 0.5X flowrate Spray 0.01X droplet size, 0.01X flowrate Spray 0.001X droplet size, 0.001X flowrate 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 Mass spilled Mass burned Combined spray and pool fire Combined spray and pool fire 2 cases not isolated

79 0

0.05 0.1 0.15 0.2 0.25 0

2000 4000 6000 8000 10000 Oxygen concnetration (-)

Time (s)

Spray base case Spray 0.5X droplet size Spray 0.5X droplet size, 0.5X flowrate Spray 0.01X droplet size, 0.01X flowrate Spray 0.001X droplet size, 0.001X flowrate 0

500 1000 1500 2000 2500 0

2000 4000 6000 8000 10000 Temperature (K)

Time (s)

Spray base case Spray 0.5X droplet size Spray 0.5X droplet size, 0.5X flowrate Spray 0.01X droplet size, 0.01X flowrate Spray 0.001X droplet size, 0.001X flowrate 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 Room temperature Oxygen concentration Combined spray and pool fire Combined spray and pool fire

80 1.E-15 1.E-14 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 0

2000 4000 6000 8000 10000 Fraction of inventory (-)

Time (sec)

Xe Cesium Ba Iodine Te Ru Mo Ce La U

Cd Ag 1.E-15 1.E-14 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 0

2000 4000 6000 8000 10000 Fraction of inventory (-)

Time (sec)

Xe Cesium Ba Iodine Te Ru Mo Ce La U

Cd Ag 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 Airborne in the sodium purification room Environmental release

81 Cover-gas pipe break scenario cv Other Reactor Building Rooms cv-42 cv-43 cv-44 cv-45 cv-46 Stack Fan HEPA Pre-filter Reactor Building HVAC Intermediate loop dump tank Printed Circuit Heat Exchanger Leakage to environment cv Containment dome Leakage to environment 110 - Cold Pool #1 cv Guard Vessel cv Reactor Cavity cv Air Gap Supply Exhaust Rails cv Sodium Purification Room Intra-building Transfer Tunnel Accident scenario assumptions

  • Cover-gas piping is not isolated
  • Discharge flow is steady and maintained by large pressure control supply tanks
  • HVAC is running with 2X air changes per hour
  • The scenario includes failure of the cladding boundary on 217 fuel rods (i.e., 1 assembly)

82 1.0E-13 1.0E-12 1.0E-11 1.0E-10 1.0E-09 1.0E-08 1.0E-07 1.0E-06 1.0E-05 1.0E-04 1.0E-03 1.0E-02 1.0E-01 1.0E+00 1

10 100 1000 10000 Fraction of inventory (-)

Time (sec)

Released In-vessel Reactor building Filters Environment 1.0E-10 1.0E-09 1.0E-08 1.0E-07 1.0E-06 1.0E-05 1.0E-04 1.0E-03 1.0E-02 1.0E-01 1.0E+00 1

10 100 1000 10000 Fraction of inventory (-)

Time (sec)

Released In-vessel Reactor building Environment Cover-gas pipe break scenario The noble gases released from the failed fuel claddings circulate with the sodium but eventually rise to the surface of the sodium pool Once in the cover gas, they leak through the cover gas pipe break.

The HVAC circulates the released gases out the plant stack The released iodine combines with sodium to form sodium iodine (NaI).

Most of the NaI remains in the pool due to its low vapor pressure in this scenario (~0.01 Pa)

The released NaI condenses into small aerosols that are not completely filtered by the HEPA Released Noble gas behavior Released NaI behavior

Examples for fuel fabrication and reprocessing safety analysis

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

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 towards the hot cells, which are the contain the processes with the highest radioactive inventories.

-0.49 kPa

-0.12 kPa 0 kPa

-0.12 kPa BNFP facility drawing - Exhaust-side ventilation 8

The processing (hot) cells are where the fire and/or explosion events are simulated. The regions are enclosed with red dotted lines The plant stack is the filtered release pathway after 2 sets of filters BNFP facility drawing 6

Hot cells were the fire and/or explosion events are simulated Hot cells for hazardous material processing Safety-grade ventilation and filtration system 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 1.E+08 1.E+09 1.E+10 1.E+11 1.E+12 1.E+13 0

3 6

9 12 Activity Release (Bq)

Time (hr)

Activity Distribution Environment Exhaust Total Hot Cells Support Gallery Other Examples of Fire Scenario Results - Radionuclide Results 12 Sensitivity of the Fire Size Modeling Accident - Key boundary conditions Fans draw released radionuclides to the filters, where they are captured by the HEPA filters The environmental release is relatively small because the HEPA filters remained intact. The activity release is due to aerosols below the min. effective HEPA capture size and radionuclide gases.

Example of an Explosion Scenario Result 14 Pressure response at the filters between the PPC and the stack 0

5 10 15 20 25 30

-5 0

5 10 15 Pressure drop (kPa)

Time (sec)

HEPA Filter 4 HEPA Filter 1 AFS VFS End of the explosion AFS & VFS always <0 psig HEPA Filter 1 fails HEPA Filter 4 fails 2.49 kPa = HEPA overpreessure failure Activity distribution in the first hour after the accident

  • Pressure response figure below shows immediate failure of HEPA Filter 7 at the exit of the PPC
  • The dissipation of the pressure from the explosion also fails the final exhaust filter within 13 seconds
  • Activity distribution above shows a large release to the environment due to the failure of the two HEPA filters between the PPC and the plant stack.
  • Pre-filter 1 remains intact and retains show larger aerosols 0%

20%

40%

60%

80%

100%

120%

0 600 1200 1800 2400 3000 3600 Activity Distribution (%)

Time (sec)

Activity Distribution Environment Total Exhaust Filter 1 Hot Cells Support Gallery Other Example of a fire scenario Example of an explosion scenario 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]

86 Reprocessing and fuel fabrication accident analysis

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

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

MELCOR Summary

88

  • 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 MELCOR SFR Summary

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

92 IAEA -TECDOC-2006 Notes No mixture compound vapor pressure Ideal mixture Raoults Law Real mixture Excess for deviation from Raoults Law CDA no mixing release insights Halogens o

NaI(l) is predominant chemical species for real mixture o

Bromine forms CsBr with 50% release at 950 K (no mixture) but drops to <10-4 with mixing Alkali metals o

Cs binds to CsI, CsRb. CsBr, CsNa with 90% release 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 No mixture assumption

93 IAEA -TECDOC-2006 Notes No mixture compound vapor pressure Ideal mixture Raoults Law Real mixture Excess for deviation from Raoults Law No mixture assumption 873 K

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