ML23058A213

HTGR Workshop 2023 Slides - Final

ML23058A213
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Issue date:	02/27/2023
From:	Shawn Campbell Office of Nuclear Regulatory Research
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SCALE & MELCOR non-LWR Fuel Cycle Demonstration Project -

High Temperature Gas-Cooled Reactors NRCs Volume 5 - Public Workshop #1 February 28, 2023 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Office of Nuclear Material Safety and Safeguards Office of Nuclear Reactor Regulations

Outline

• NRC Strategy for non-LWRs Readiness

• Project Scope

• HTGR Nuclear Fuel Cycle

• Overview of the Simulated Accidents

• Nuclide inventory, decay heat, and criticality calculations in SCALE

• High-Temperature Gas-Cooled Reactor Modeling using MELCOR

• Summary & Closing Thoughts

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

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

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 (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

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

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

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

Representative Fuel Cycle Designs

• Completed 5 non-LWR fuel cycle designs for -

• HPR - INL Design A

• HTGR - PBMR-400

• FHR - UCB Mark 1

• MSR - MSRE

• SFR - 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

Overview of the HTGR fuel cycle SCALE/MELCOR Non-Light-Water Reactor (Non-LWR)

Fuel Cycle Demonstration Project for a High Temperature Gas-Cooled Reactor NRC Public Workshop, February 28, 2023 F. Bostelmann, E. Davidson, W. Wieselquist 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 effort was to identify hazards across the HTGR fuel cycle

• Determine details of the fuel cycle stage based on publicly available information Use PBMR-400 as basis for fuel pebble details and for HTGR operation Identify where additional data are needed or can benefit simulations

• Identify potential hazards and accident scenarios for each stage of the fuel cycle Identify accidents independently of their probability to occur Select accident scenarios for SCALE/MELCOR to simulate under consideration of the project goal to demonstrate SCALE/MELCORs capabilities

• Challenges encountered during the scenario development

• Some stages of the HTGR fuel cycle are not yet developed

• Many documents are proprietary 11

HTGR Fuel Cycle

- Enrichment - Transportation of UF6 - TRISO fabrication - Pebble fabrication

- Gas centrifuges E1 - Package: DN30-X T1 - Sol-gel process F1 - Process: X-energy F2

- Transportation of - Discharged pebble

- Fresh fuel staging - Power production storage onsite fuel pebbles

- Package: Versa-Pac T2 - Pebble loading U1 - Online refueling U2 - Spent fuel/used fuel/graphite tanks U4 Not covered: uranium mining & milling, spent fuel transportation & off-site storage & final disposal 12

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, initially 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 13

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

License application under review by NRC 30B-X cylinder similar to 30B cylinder, but with criticality control system (addition of internal absorber structure) 30B cylinder: Licensed up to 5 wt.% 235U; permissible UF6 mass of 2277 kg Permissible mass in DN30-X depends on enrichment (proposed):

Package design Enrichment limit Permissible UF6 mass DN30-10 10 wt.% 235U 1460 kg DN30-20 20 wt.% 235U 1271 kg 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:

ORANO: 30B cylinder with DN30 PSP

• Criticality due to water accidents and container drop

• Release of UF6 due to container rupture Ref.: ORANO Safety Analysis Report for the DN30-X Package https://www.nrc.gov/docs/ML2232/ML22327A183.pdf 14

F1: Fabrication of TRISO Particles Fuel kernel:

• U.S. TRISO production based on internal sol-gel process

• Starting sol is a uranyl nitrate solution

• Sol is dripped through a nozzle into a heated organic diluent (silicone oil)

• Heat causes HMTA (Hexamethylenetetramine) to chemically decompose and induces a gelation reaction which eventually forms the fuel kernel Kernel coating:

Ref.: P. Pappano, TRISO-X Fuel Fabrication Facility Overview, Introductory Meeting with

• Coat the kernels with the carbon layers using the NRC, ML18254A086 (2018). https://www.nrc.gov/docs/ML1825/ML18254A086.pdf various gas mixtures at different temperatures Major hazards:

• Hazards from the use of the various chemicals (spills, reaction with water, fire, explosion)

• Criticality due to improper storage of UF6 or water accidents Ref.: R. L. Seibert, et al., Production and characterization of TRISO fuel particles with multilayered SiC, Journal of Nuclear Materials, 515, pp. 215-226 (2019).

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F2: Fabrication of Fuel Pebbles

• Graphite powder is dried, pulverized and then is used for overcoating the TRISO kernels at controlled temperatures

• Pre-press overcoated TRISOs onto inner graphite sphere

• Final pressing of entire pebble which includes outer non-fuel region followed by some steps before pebble is released for inspection1 PBMR-400 fuel pebble and TRISO particle2 Major hazards:

• Criticality due to improper storage of TRISOs or fuel pebbles Ref 1: IAEA, Fuel performance and fission product behavior, IAEA-TECDOC-978 (1997).

• Contact with water leading to graphite corrosion Ref 2: "PBMR Coupled Neutronics / Thermal-hydraulics Transient Benchmark The PBMR-400 Core Design, Vol. 1: The Benchmark

• Development of graphite dust leading to fire hazard Definition," NEA/NSC/DOC(2013)10, 2013.

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T2: Transportation of Fresh Fuel Pebbles Versa-Pac:

• Package for shipping of of fuel pebbles and storage at the plant

• Versa-Pac is licensed for enrichments up to 100% 235U

• Maximum allowed mass determined by enrichment:

584 PBMR-400 pebbles with 9.6 wt% 235U enrichment Major hazards: Versa-Pac

• Criticality due to water accidents and container drop

• Contact with water leading to graphite corrosion Ref.: DAHER-TLI Versa-Pac Safety Analysis Report

• Development of graphite dust leading to fire hazard https://www.nrc.gov/docs/ML1833/ML18330A093.pdf 17

U1/U2/U4 - Utilization Stages

Reference:

PBMR-400

• Daily fuel pebble circulation: 2,900 pebbles

• Average number of passes per fuel pebble: 6

• Number of fresh fuel pebbles loaded per day: 483

• 25 fuel pebble canisters per month if canister loaded to 235U limit

• 40 VP-55 canisters per month according to our model (see SCALE slides)

• Plant lifetime: 40 years

• Total number of fuel pebbles during lifetime, considering 6 overhauls: 6,969,667

• Target burnup: 90 GWd/tHM

• Fuel enrichment: 9.6 wt% 235U

• Total pebble loading in core: 451,530 pebbles (start-up core: 2/3 graphite pebbles)

• Pebble handling via Fuel Handling and Storage System (FHSS)

PBMR-400 Ref.: PBMR Coupled Neutronics/Thermal-hydraulics Transient Benchmark, The PBMR-400 Core Design -

Volume 1: The Benchmark Definition. Technical Report NEA/NSC/DOC(2013)10, OECD/NEA, 2013. 18

U1: Fresh Fuel Staging and Loading

• Fresh pebbles stored in Versa-Pac containers

• Pebbles are fed into system via hopper(s)

• Pebbles enter the fuel handling and storage system one by one

• Also consider graphite pebbles for startup core Major hazards:

• Criticality due to water accidents, graphite pebble misloading, tank rupture

• Development of graphite dust leading to fire hazard 19

U2: Power Production Including Online Refueling Fuel Handling and Storage System:

• Loading and unloading of pebbles into and from the reactor core while the reactor is operating at power

• Integrity verification: Separate out broken/damaged spheres

• Measurement of each fuel pebbles burnup via gamma spectroscopy

• Lift the sphere to the top of the reactor through pneumatic pressure tubes and other means Major hazards:

• Criticality due to pebble misloading, incorrect burnup measurement, failed core unloading device

• Temperature increase in pipes or core due to stuck pebbles Ref.: C. C. Stoker et al. PBMR Used Fuel

• Fission product release into coolant or adsorption into graphite dust Storage Criticality for Most Reactive Core Loading. Proc. ICNC, St. Petersburg, Russia, 28

• Graphite oxidation due to chemical attack May-1June, pages 8-14, 2007.

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U4: Onsite Discharged Pebble Storage Ref.: J. Slabber. Reactor Unit and Main Support Systems.

• 10 Spent Fuel Tank (SFT): https://www.nrc.gov/docs/ML0 606/ML060680079.pdf, 2006

• 620,000 pebbles per container

• Interim storage of up to 80 years (40 years of reactor operation + 40 years of additional onsite storage)

• 1 Graphite Storage Tank (GST)

• Graphite pebbles from startup core

• 1 Used Fuel Tank (UFT):

• unloading of pebbles from core for maintenance, reflector replacement etc.

Major hazards:

• Criticality due to water accidents, graphite pebble misloading, tank rupture

• Insufficient heat removal due to failed cooling

• Release of fission products from damaged pebbles Ref.: C. C. Stoker et al. PBMR Used Fuel Storage Criticality

• Development of graphite dust leading to fire hazard for Most Reactive Core Loading. Proc. ICNC, St. Petersburg, Russia, 28 May-1June, pages 8-14, 2007.

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

• Use of High Assay Low Enriched Uranium (HALEU) fuel with up to 19.75 wt% 235U

• No approved commercial size transportation and storage packages for UF6 and fresh fuel pebbles

• New chemicals and processes for TRISO particle and fuel pebble fabrication

• Continuous circulation of fuel pebbles with removal of depleted pebbles during operation

• Handling of irradiated fuel pebbles during operation Major identified hazards:

• Higher enrichment impacting criticality during UF6 and fuel pebble storage and transportation

• Hazards from the use of the various chemicals (spills, reaction with water, fire, explosion)

• Graphite corrosion leading to fuel pebble damage, and graphite dust leading to fire hazard

• Fission product release from damaged fuel pebbles Additional details needed:

• Onsite fresh fuel pebble and graphite pebbles storage details

• Fuel pebble handling and (un)loading procedure (pressure boundaries, canisters, loading devices, etc.)

• Onsite spent fuel pebble storage design details

• HTGR containment and building design details 22

Nuclide inventory, decay heat, and criticality calculations with SCALE for the HTGR fuel cycle R. ELZOHERY, D. HARTANTO, F. BOSTELMANN, W. WIESELQUIST NRC PUBLIC WORKSHOP FEBRUARY 28, 2023 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 SCALE capabilities for simulating different stages of the HTGR fuel cycle

• Selected scenarios for demonstration

• UF6 transportation Scenario 1: Water ingress into array of canisters at optimal moderator to fuel ratio Analysis: Perform SCALE criticality calculations*

• Fresh fuel pebble transportation Scenario 2: Damage/drop of a container leading to reduced array spacing and potential criticality Analysis: Perform SCALE criticality calculations*

• Fuel utilization Scenario 3: FHSS pipe rupture: pebbles exit out of the reactor with high temperature and pressure, leading to graphite and air interaction Analysis: Determine equilibrium core, simulate individual pebbles; MELCOR selects target pebbles for severe accident progression

• Onsite storage of spent fuel Scenario 4: Collision of vehicle or suspended load with storage tank causing damage to tank and damage to pebbles inside tank, causing fission product and graphite dust release Analysis: Use individual pebbles to build up inventory in a storage tank; MELCOR uses tank decay heat/inventory for severe accident progression

• This is not full certification type analysis, but an analysis for demonstration of capabilities 24

Reference HTGR: PBMR-400 PBMR-400 TRISO particle and fuel pebble [1]

Characteristic Value Thermal power 400 MWth Fuel enrichment 9.6 wt.% 235U Target discharge burnup 90 GWd/MTU Number of pebbles in core ~452,000 PBMR-400 SCALE model [2]

[1] Nuclear Science Committee, Nuclear Energy Agency (NEA), "PBMR Coupled Neutronics / Thermal-hydraulics Transient Benchmark The PBMR-400 Core Design, vol. 1: The Benchmark Definition," NEA/NSC/DOC(2013)10, Paris, France, 2013.

[2] S. E. Skutnik and W. A. Wieselquist, "Assessment of ORIGEN Reactor Library Development for Pebble-Bed Reactors Based on the PBMR-400 Benchmark", ORNL/TM-2020/1886, Oak Ridge National Laboratory, Oak Ridge, TN (July 2021) 25

Scenario 1: Water ingress into packages during UF6 transportation 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

DN30-X UF 6 transportation package

• DN30-X is a new transportation package with neutron poisons designed for HALEU

• X is a specific design identifier. Either 10 for a maximum enrichment of 10 wt% or 20 for a maximum enrichment of 20 wt%

• The package contains:

• Protective Structural Packaging (PSP)

• 30B-X cylinder (30B-10 or 30B-20)

• Both 30B-10 and 30B-20 have identical outer dimensions to the standard 30B cylinder

• The 30B-X cylinder contain Criticality Control Rods (CCRs) DN30-X package*

made of boron-carbide

• The PBMR fuel enrichment, 9.6 wt% 235U, is used for calculations with the 30B-10 model

• The maximum HALEU enrichment, 20 wt%, is used for calculations with the 30B-20 model

• Safety Analysis Report for the DN30-X Package https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML22327A183 30B-X cylinder* 27

SCALE model of DN30-X package Model tools and data:

• Neutron transport code: SCALEs Monte Carlo code Shift

• Shift is optimized for performance in parallel fast results with multiple cores Control rod

• keff calculations converged to 10 pcm statistical uncertainty UF6

• Nuclear data versions: ENDF/B-VII.1 and ENDF/B-VIII.0 PSP continuous energy libraries Conservative modeling assumptions:*

• Lattice holder, valve, plug, and nameplate are neglected

• The foam material in the PSP is neglected

• UF6 is assumed at a theoretical density of 5.5 g/cm3 with 0.5 wt %

HF impurities

• Cylinders are 100% filled with UF6 (exceeds the permissible mass for the 30B-10 and 30B-20 cylinder; this is conservative from criticality safety perspective)

• The same assumptions in the safety analysis report are adopted: Safety Analysis Report for the DN30-X 3D view of SCALE model Package, https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML22327A183 28

SCALE baseline result for DN30-X Nuclear Data Library DN30-10 DN30-20 keff ENDF/B-VII.1 CE 0.58459 +/- 0.00011 0.77772 +/- 0.00011 keff ENDF/B-VIII.0 CE 0.58549 +/- 0.00010 0.77761 +/- 0.00011 k (pcm) 90 +/- 15 -11 +/- 16 Control rod UF6 Infinite hexagonal PSP array of packages touching on sides, surrounded by airno water ingress.

30B-10 30B-20 (33 control rods) (43 control rods)

• The same assumptions in the safety analysis report are adopted: Safety Analysis Report for the DN30-X 3D view of SCALE model*

Package, https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML22327A183 29

Impact of water on criticality for DN30-10

• Minimum package pitch (touching) is the most Water reactive configuration.

surrounding

• Water ingress into PSP Water and inside has lower keff.

surrounding PSP PSP 30

Impact of water on criticality for DN30-20 DN30-20 shows the same Water trends as DN30-10.

surrounding Water PSP surrounding Additional moderation from and inside surrounding water or ingress PSP into PSP, decreases keff from baseline.

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Scenario 2: Damage/drop during fresh fuel pebble transportation 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

Fresh pebble transportation package

• Versa-Pac Package:

• 55-gallon package (VP-55)

• The payload containment area is contained in a drum for enhanced structural protection.

• The packages interior is completely insulated with the appropriate layers of ceramic fiber.

• Mass loading of 235U is determined by enrichment.

• Fuel pebbles:

• PBMR-400 fuel pebbles Characteristic Value Fuel enrichment 9.6 wt.% 235U TRISO packing fraction ~9%

Uranium content per pebble 9g

• Container permits maximum of 584 pebbles based on given enrichment, and up to 505 g of 235U permitted loading. 364 pebbles fit into container Versa-Pac Illustration*

at 55% packing fraction.

• Versa-Pac Safety Analysis Report, https://www.nrc.gov/docs/ML1833/ML18330A093.pdf 33

SCALE model of the VP-55 Model tools and data:

• Neutron transport code: SCALEs Monte Carlo code Shift

• Shift is optimized for performance in parallel fast results with multiple cores

• keff calculations converged to 10 pcm statistical uncertainty

• Nuclear data versions: ENDF/B-VII.1 and ENDF/B-VIII.0 continuous energy and multi-group libraries

• Continuous-energy model: TRISO particles are explicitly modeled and randomly distributed inside the fuel sphere

• Multi-group model: TRISO particles in fuel sphere modeled via double-heterogeneous unit cell for resonance treatment Model details:

• 364 pebbles are placed inside the container, equivalent to 315 grams of 235U, and 55% packing fraction (assumption)

• Reflective boundary conditions account for an array of containers

• Insulation specifications are not well-defined, since they depend on the manufactures and fabrication, but the used material densities are within the 3D SCALE VP-55 model recommended limits 34

SCALE baseline result for the VP-55

Reference:

Infinite array of touching containers surrounded by air library keff +/- sigma kMG-CE (pcm) kENDF (pcm)

ENDF/B-VII.1 CE 0.30387 +/- 0.00010 (ref) (ref)

ENDF/B-VII.1 252g 0.30416 +/- 0.00010 29 +/- 14 ENDF/B-VIII.0 CE 0.30575 +/- 0.00010 (ref) 188 +/- 14 ENDF/B-VIII.0 252g 0.30486 +/- 0.00010 90 +/- 14

• Runtime comparison:

• SCALE 6.3: CE runtime 20x MG runtime

• SCALE 7.0 development: CE runtime 2x MG runtime

• Impact of fuel pebble random distribution:

• Mean of bias and bias uncertainty due to random pebbles distribution is studied by running 10 different random realizations with ENDF/B-VII.1 252g

• Average keff: 0.30406 +/- 0.00003

• Difference to reference result: keff = 10 +/- 10 (pcm)

The impact of the explicit pebble distribution in this model is negligible 35

Impact of damage/drop on criticality for VP-55 PF = 0.55 (364 pebbles) PF = 0.60 (397 pebbles)

Potential increase in pebble packing fraction and package array spacing

• Both packing fraction and package array spacing increase keff from baseline to a slight optimum at 14-16 cm spacing.

• Max potential increase ~300pcm for the PF=0.6 case.

• Array of packages still very low with max keff ~0.33.

Additional water/flooding scenarios for VP-55 b) Impact of water surrounding containers a) Impact of water surrounding the containers and water ingress into the container For all cases, largest keff found when cylinders are touching (unlike the air-only case)

Scenario 3: Pebble ejection from fuel handling system Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energys National Nuclear Security Administration under contract DE-NA0003525. SAND20XX-XXXX P

SCALE approach for fuel inventory generation

• Zone-wise equilibrium core inventory:

• The SCALE PBMR-400 core model1 was divided into 5 radial channels and 22 axial regions

• Zone-average inventory corresponding to an equilibrium state was generated with an established approach2

• Core-average inventory is equal to the inventory of a used fuel tank (UFT) which contains all pebbles during maintenance

• An inventory interface file with core-average inventory was provided to MELCOR

• Rapid inventory of 20,000 individual pebbles:

• Inventory was generated based on random pebbles histories, considering different radial channels and associated power distributions3

• Seven passes were simulated for each pebble

• An inventory interface file containing the 20,000 pebble inventories was provided to MELCOR

[1] S. E. Skutnik et al. (2021), ORNL/TM-2020/1886, Oak Ridge National Laboratory, Oak Ridge, TN SCALE PBMR-400 model

[2] F. Bostelmann, et al. (2021), ORNL/TM-2021/2273, Oak Ridge National Laboratory, Oak Ridge, TN

[3] D. Hartanto, et al. (2022),"Uncertainty Quantification of Pebble's Discharge Burnup and Isotopic Inventory Using SCALE," Proc. ANS Winter Meeting, Phoenix, AZ, November 13-17.

39

Characteristics of pebbles in PBMR-400 Pebble burnup [GWd/MTU]

Average pebble burnup after each pass Burnup distribution after each pass Target burnup is 90 GWd/MTU, but 7 passes are simulated to The error bars correspond to the burnup range after each pass include pebbles that havent reached the target burnup at 6th pass.

Target burnup and number of passes

• A pebble is retired earlier than the target burnup in case it has a chance to exceed the target if it is returned to the core.

• A burnup cutoff has to be chosen after which pebbles are removed from the system Selected limit

• With burnup limit at BUMS of 85 GWd/MTU, the average

• Fraction of retired pebbles using cut-off value 85 GWd/MTU burnup of the retired pebbles is 90 GWd/MTU (target

• On average, it takes 6 passes through the core for a pebble burnup). to reach the target burnup of 90 GWd/tU.

Average decay heat of PBMR pebbles Average decay heat of a pebble at the end of Pebble power decrease with passes leading to a each pass decrease in the decay heat 42

Top contributors to the decay heat of PBMR pebbles Pass 1 Pass 3 Pass 6 Fission products dominate in early passes because of higher fission rate, then actinides begin to appear among the top 5 contributors in late passes 43

Scenario 4: Collision with the spent fuel storage tank 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

PBMR-400 spent fuel tank

• After a pebble is retired, the FHSS moves the pebbles to the spent fuel tank (SFT)

• One SFT can store 620,000 pebbles

• The PBMR-400 has multiple SFTs that together can store all pebbles from entire reactor lifetime

• Interim storage of up to 80 years (40 years of operation + 40 years of onsite storage)

• 483 fuel pebbles are discharged daily

• It takes ~1,284 days to fill one tank PBMR-400 FHSS with pebble storage tanks*

• J. Slabber. Reactor Unit and Main Support Systems.

https://www.nrc.gov/docs/ML0606/ML060680079.pdf, 2006 45

SFT modeling procedure

• The SFT is filled one day at a time in 1,284 layers

• The discharge inventory of the 20,000 pebbles is blended to compute average discharge inventory.

• Each layer is decayed to the time when the tank is full, as shown on the right.

• An interface inventory file containing inventory of each slice in the spent fuel tank is provided to MELCOR team for accident analysis Spent fuel tank 46

Total decay heat of spent fuel tank (620k pebbles) 47

Total decay heat of spent fuel tank (620k pebbles)

Top Layer Decay Heat The top layers are dominating decay heat and the sharp Top Layer drop is driven by nuclides in Contributors that range.

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SCALE 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

Summary of HTGR fuel cycle hazard analysis - SCALE criticality, nuclide inventory, and decay heat

1. Water ingress into DN30-X UF6 transportation packages

• With additional neutron absorbers, baseline infinite array of packages significantly subcritical, max keff ~

0.78, even for 20 wt.% 235U enr.

• keff still shows large margin to criticality with any amount of water ingress

2. Damage/drop of a VP-55 fresh fuel pebble transportation package

• Small package with 350-400 pebbles per package

• Using PBMR-400 pebbles with ~10 wt.% enr., keff ~0.3; for 20 wt.% enr. keff ~0.5

• Strong impact of pebble packing fraction: 2,000 pcm increase with 5% packing fraction increase

3. Pipe rupture in FHSS

• 20,000 pebbles were simulated to yield variations in inventory/decay heat

• Actual accident progression to be handled by MELCOR using SCALE inventory data

4. Damage to SFT potentially causing loss of cooling and/or fission product release

• SFT inventory/decay heat generated using 20,000 pebble histories

• Actual accident progression to be handled by MELCOR using SCALE inventory data 50

Conclusions of SCALE analysis

• SCALE capabilities to simulate different scenarios in different fuel cycle stages were demonstrated.

• Analysis involved criticality calculations, fuel inventory and decay heat calculation, and radionuclide characterization. Results obtained are physically reasonable and follow expectations.

• SCALE has been well validated for criticality and reactor fuel depletion of water-moderated LEU systems*. Additional benchmarks are needed to extend validation to graphite-moderated and HALEU systems.

• Additional information is needed for improved analysis: commercial size transportation canisters for UF6 and fuel pebbles, handling of fuel pebbles during operation (addition of fuel pebbles to the FHSS, extraction of fuel pebbles, etc.),

onsite storage of spent fuel pebbles, etc.

• See SCALE validation reports: https://www.ornl.gov/scale 51

High-Temperature Gas-Cooled Reactor Modeling using MELCOR Lucas I. Albright, Kenneth Wagner, David L. Luxat SAND2023-12955PE

MELCOR HTGR Fuel Cycle Modeling Fission product release and transport

• Release from TRISO kernel

• Radionuclide distributions within the layers in the TRISO particle and compact

• Release to coolant Hazardous material release and transport

• U-bearing materials Slabber (2006), Technical Description of the PBMR Demonstration

• Corrosives Power Plant, https://www.nrc.gov/docs/ML0609/ML060940293.pdf Other phenomenological models

• Graphite oxidation

• Intercell and intracell conduction Fuls, W.F., Mathews,

• Convection & flow E.H. (2006). Passive Cooling of the PBMR

• Control function-based chemistry spent and used fuel tanks. Nuclear Engineering and Design, 237, 1354-1362.

53

Capability Demonstration

• The modeled systems and results are representative of prototypic HTGR fuel cycle systems and postulated accidents.

• The modeled systems have been derived from conceptual designs

• The calculations are intended to illustrate modeling capabilities

• No safety judgments should be concluded 54

MELCOR Models and Simulations A Short Summary of Facility Modeling with MELCOR Source term and leak path factor analysis (aerosol physics, vapor physics, user-defined speciation and chemistry, etc)

Images of the title pages for - Broad accident sequence spectrum (multi-room

• barnwell fire, explosions, spills, etc.)

• nsrd-10 Complex facility modeling (connectivity, interlocks, multi-zone ventilation and filtration, etc.)

• MELCOR capabilities facilitate radiological and non-radiological hazard analyses 56

Facility Demonstration Facility Model Demonstration facility overview with relative locations of fuel storage Slabber (2006), Technical Description of the PBMR Demonstration Power Plant, https://www.nrc.gov/docs/ML0609/ML060940293.pdf tank cubicle and UF6 cylinder storage featured in the following slides

• Generator housing compartment (GHC)

• Reactor crane compartment (RCC)

• Power conversion compartment (PCC)

• Storage compartment (SC)

• Power conversion crane compartment (PCCC)

• Auxiliary Compartment (AC)

• Reactor compartment (RC) 58

Facility Model Detail Compartment Volume [m3]

90 Environment 1000.0 Intake 1000.0 61 Stairwell 3200.0 Auxiliary Auxiliary 12400.0 Altitude [m] (not to scale)

Exhaust Compartment Intake Compartment Storage 38400.0 Environment Compartment Building Filter 10.0 30 Stairwell 26 Building Exhaust 1000.0 Filter Storage 23 Supply Flow Exhaust Flow 4 Compartment Doors

• Flow connections not 0 representative of connection

-2 altitudes

-10 59

Building Filter Detail 30 Compartment Volume [m3]

Pre-filter 5.0 To Building Exhaust HEPA 2.0 Altitude [m] (not to scale)

Fan Inlet 2.0 Fan Outlet 2.5 From Building Fan Outlet Compartments Specification -

Fan P [Pa] 100.0 26 Supply Flow Fan Inlet HEPA Filter Pre-filter Exhaust Flow Fan

• Flow connections not representative of connection altitudes 23 60

UF 6 Cylinder E1: Enrichment - UF 6 Cylinder Rupture Scenario Summary

• Overfilled model 48Y cylinder is heated resulting in tank rupture and UF6 release as vapor and aerosol NUREG/CR-6410 Scenario 6 - Case 1 (based on NUREG-1179)

• Rapid and complete release of massive quantity of UF6 Flashing ratio = 0.45 vapor and 0.55 solid particles UF6+2H2O UO2F2 + 4HF + 117.147 kJ/(kg mol UF6)

Demonstration Characteristics and Important Phenomena

• MELCOR modeling flexibility (reproduction of NUREG/CR 6410 analysis w/ MELCOR)

• Aerosol and vapor RN sources after tank rupture

• Material transport by and NCG/CVH package

• Material transport by RN package

• Control function-based species chemistry 62

Demonstration UF 6 Cylinder U.S. Nuclear Regulatory Commission (1986). Rupture of Model 48 UF6 Cylinder and Release of Uranium Hexafluoride. NUREG-1179, Volume 1, U.S Nuclear Regulatory Commission 63

UF 6 Cylinder Detail Compartment Volume [m3]

90 UF6 Storage 1000.0 Specification -

UF6 release mass 14000 61 [kg]

Flashing Ratio 0.45 vapor/0.55 aerosol Altitude [m] (not to scale)

Building Relative 0.4 Humidity Release Duration [s] 1.0 x 10-3 30 Door Open Fraction 1.0 26 23 Supply Flow Exhaust Flow 4 Doors UF6 Storage *Flow connections not representative 0 of connection altitudes

-2

-10

UF6 Cylinder - Catastrophic Rupture Building Flow Building Pressure Building Reaction Heat Temperature Generation Rate UF6+2H2O UO2F2 + 4HF + Q

• Rupture event causes a large pressure spike and mass ejection to atmosphere through building openings

• Elevated building temperatures are observed after the rupture and are sustained by exothermic reactions 65

UF6 Cylinder - Catastrophic Rupture Continued UF6+2H2O UO2F2 + 4HF + Q Material Type Material Material Species Transport UF6 Transport UO2F2 HF Transport Transport

• U-bearing mass released primarily during initial rupture event, minimal releases observed thereafter

• U-bearing masses are primarily aerosol and exhibit strong tendency to deposit on building structures 66

UF 6 Cylinder Sensitivities UF6 Cylinder Sensitivity Specification Model Parameter Distribution Range Vapor Fraction uniform 0.0 - 1.0 Model Release Duration log-uniform 1.0e 600.0 Parameters UF6 Storage Door Area Multiplier uniform 0.01 - 1.0 Relative Humidity uniform 0.01 - 0.99 Quantities Of Interest Reaction Heat Generation Released U-bearing Mass Filtered U-bearing Mass Released HF Mass Rate

Model Sensitivities to Peak Quantities of Interest Reaction Heat Generation Released U-bearing Mass Filtered U-bearing Mass Released HF Mass Rate

• No quantities of interest exhibit notable correlation to the door open area fraction

• Vapor fraction exhibits a strong, positive correlation to quantities of interest

• Relative humidity exhibits a strong impact on quantities of interest

• Weaker negative correlation to release duration is exhibited for quantities of interest for release durations <100s, correlation strength increases for release durations >100s

Fuel Storage Tank 70

Demonstration Fuel Storage Inlet Spent Fuel: Retired fuel that has reached a specified Exhaust burnup and cannot be reloaded into the core Used Fuel: Fuel that has not reached the specified Concrete Cubicle burnup and can be reloaded into the core Gamma Shield

• May require temporary storage during core maintenance Thermal Shield Downcomer Storage Tank Cooling Tubes Fuel Spheres MELCOR fuel storage tank concept overview with Fuls, W.F., Mathews, E.H. (2006). Passive Cooling of the PBMR spent and used designated coolant flow fuel tanks. Nuclear Engineering and Design, 237, 1354-1362. 71

Demonstration Fuel Storage: Operational Modes Closed Loop Active Cooling Open Loop Active Cooling Open Loop Passive Cooling

• Normal operational mode for

• Normal operational mode for

• On loss of power, louvres spent fuel storage tanks used fuel storage tanks open (transition from closed to

• nominal decay heat ~140kW

• nominal decay heat ~640kW open loop) and/or active

• Building flow is isolated from

• Concrete cubicle draws on cooling is lost (spent or used concrete cubicle flow building air supply fuel, respectively)

MELCOR fuel storage tank operational mode concept overview with designated coolant flow 72

Cubicle Model Additions Supply Flow 90 Exhaust Flow Doors Rupture 61 *Flow connections not representative of connection altitudes Altitude [m] (not to scale) 30 26 Cubicle Filter 23 Fuel Storage 4 Cubicle 0

-2

-10 73

Fuel Storage Cubicle Detail Cubicle 90 Filter Cubicle Cubicle 61 Inlet Exhaust Storage Compartment Altitude [m] (not to scale)

Plenum Reference Diagram Top with design flow Storage Tank 30 Downcomer Bypass 26 Middle Closed Loop Flow 23 Open Loop Flow Rupture Lower 4 *Flow connections not Bottom representative of 0 connection altitudes

-2 74

Fuel Storage Tank Detail Fuel storage tank axial nodalization Fuel storage tank radial nodalization Compartment Volume [m3]

Concrete Cubicle 800.0 Fuel Storage Tank 70.0 Specification -

Used Fuel cubicle fan P with filter [Pa] 2000.0 Used Fuel cubicle fan P without filter [Pa] 100.0 Spent Fuel Fan P [Pa] 10.0 Heat Exchanger Power Logarithmic Mean Temperature Fuls, W.F., Viljoen, C., Stoker, C., Koch, C., Kleingeld, Difference M. (2005). The interim fuel storage facility of the PBMR.

Annals of Nuclear Energy, 32, 1854-1866. 75

Cubicle Filter Detail 30 Compartment Volume [m3]

Pre-filter 5.0 HEPA 2.0 To Building Exhaust Carbon 2.0 Altitude [m] (not to scale)

Fan Inlet 2.0 Fan Outlet 2.5 From Cubicle Exhaust Fan Outlet Specification -

Fan P [Pa] 100.0 26 Carbon Filter Supply Flow Fan Inlet HEPA Filter Pre-filter Exhaust Flow Fan

• Flow connections not representative of connection altitudes 23 76

Postulated Scenario Event Tree

• MELCOR flexibility facilitates exploration of large event spaces 77

Used Fuel Storage Tank 78

U2: Utilization/Online Refueling - Used Fuel Storage Tank Transients Scenario Summaries

• Normal operations - open loop active cooling

• Spurious loop closure - transition from open loop active cooling to closed loop active cooling resulting in limited airflow through the used fuel cubicle and subsequent heatup

• Loss of power - transition from open loop active cooling to open loop passive cooling resulting in reduced airflow through the used fuel cubicle and subsequent heatup

• Sensitivities without cubicle filtration - smaller fans can be used to develop similar cubicle flows when there is not a cubicle filtration system (system description does not indicate presence of filtration system)

Demonstration Characteristics and Important Phenomena

• Fuel radionuclide inventory development using SCALE

• TRISO modeling for non-reactor geometries

• Thermal hydraulics

• Used fuel storage tank operational modes and transients

• RN release and subsequent RN transport 79

Fuel Storage Tank - Used Fuel w/ Active Open Loop Heat Removal w/out Cubicle Filtration Forced convection Fuel Temperatures Decay Power Cubicle Temperatures Cubicle to Building Flow

• Normal operations exhibit decreasing fuel and cubicle temperatures as short-lived isotopes decay

• Without filtration, a smaller fan (100.0 Pa P) is needed to adequately cool the fuel and storage cubicle 80

Fuel Storage Tank - Used Fuel w/ Active Open Loop Heat Removal w/ Cubicle Filtration Forced convection Fuel Temperatures Decay Power Cubicle Temperatures Cubicle to Building Flow

• Normal operations exhibit decreasing fuel and cubicle temperatures as short-lived isotopes decay

• With filtration, a larger fan (2000.0 Pa P) is needed to adequately cool the fuel and storage cubicle 81

Fuel Storage Tank - Used Fuel w/ Spurious Loop Closure w/out Cubicle Filtration Forced convection Forced and active convection cooling Fuel Cubicle Cooling Temperatures Cubicle Cubicle Flow Temperatures

• When the cubicle does not have a filtration system, the smaller fan does not provide adequate cooling of the fuel and storage cubicle under a spurious loop closure 82

Fuel Storage Tank - Used Fuel w/ Spurious Loop Closure w/ Cubicle Filtration Forced Fuel convection Temperatures and active Forced convection cooling Cubicle Cooling Cubicle Cubicle Flow Temperatures

• When the cubicle does have a filtration system, the larger fan provides significant mass flow and adequate cooling of the fuel and storage cubicle under a spurious loop closure 83

Fuel Storage Tank - Used Fuel w/ Loss of Power w/out Cubicle Filtration Natural convection Fuel Temperatures Building Exhaust Flow Cubicle Temperatures Cubicle to Building Flow

• The unobstructed path from the cubicle exhaust to the building exhaust (i.e., no cubicle filtration) facilitates production of a natural convection loop

• Maintains adequate cooling of the fuel and storage cubicle 84

Fuel Storage Tank - Used Fuel w/ Loss of Power w/ Cubicle Filtration Natural convection Fuel Building Exhaust Flow Temperatures Cubicle to Building Flow Cubicle Temperatures

• The tortuous path of the cubicle filtration system obstructs production of a natural convection loop

• Cannot maintain adequate cooling of the fuel and storage cubicle 85

Spent Fuel Storage Tank 86

U4: Discharged Pebble Storage - Spent Fuel Storage Tank Transients Scenario Summaries

• Normal operations - closed loop active cooling

• Storage Tank and/or Cubicle Rupture - rupture configurations that allow disruption of cubicle cooling and/or release of fission products

• Loss of Forced Flow and/or Active Cooling - Loss of cubicle cooling systems causing disruption of cubicle cooling

• Loss of power - transition from closed loop active cooling to open loop passive cooling resulting different airflow through the spent fuel cubicle

• Loss of power with storage tank and cubicle rupture - transition from closed loop active cooling to open loop passive cooling resulting different airflow through the spent fuel cubicle Demonstration Characteristics and Important Phenomena

• Spent fuel radionuclide inventory development using SCALE

• Fuel modeling for non-reactor geometries

• Thermal Hydraulics

• Spent fuel Fuel storage tank operational modes and transients

• RN release and subsequent RN transport

• Graphite oxidation 87

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal Forced convection and active cooling Fuel Temperatures Decay Power Cubicle Temperatures Cubicle Flow

• Normal operations exhibit decreasing fuel and cubicle temperatures as short-lived isotopes decay

• Even with filtration, only a small fan (10.0 Pa P) is needed to adequately cool the fuel and storage cubicle 88

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Tank Rupture Forced convection and active cooling Rupture Site Flow Fuel Temperatures Cubicle Temperatures Cubicle Flow

• Spent fuel storage tank is robust to a tank breach

• Adequate cooling of the fuel and storage cubicle is maintained 89

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Cubicle Rupture Forced convection and active cooling Rupture Site Flow Fuel Temperatures Cubicle Temperatures Cubicle Flow

• Spent fuel storage tank is robust to a cubicle breach

• Forced convection maintains adequate cooling of the fuel and storage cubicle even with the rupture 90

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Tank and Cubicle Rupture Forced convection Rupture Site and active cooling Flow Fuel Temperatures Cubicle Temperatures Cubicle Flow

• Spent fuel storage tank is robust to a combined tank and cubicle breach.

• Forced convection maintains adequate cooling of the fuel and storage cubicle even with the ruptures 91

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Fan Failure Natural convection and active cooling Fuel Temperatures Cubicle Cooling Cubicle Temperatures Cubicle Flow

• Spent fuel storage tank is robust to loss of forced convection

• Natural convection is established and maintains adequate cooling of the fuel and storage cubicle 92

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Heat Exchanger Failure Forced convection without heat removal Cubicle Cooling Fuel Temperatures Cubicle Cubicle Flow Temperatures

• Spent fuel storage tank is challenged by loss of active cooling

• Without active cooling, the fuel and cubicle atmosphere heats up 93

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Fan and Heat Exchanger Failure Natural convection without heat removal Cubicle Cooling Fuel Temperatures Cubicle Cubicle Flow Temperatures

• Spent fuel storage tank is challenged by combined loss of forced convection and active cooling

• Without active cooling, the fuel and cubicle heat up in similar form to isolated loss of active cooling 94

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Loss of Power Forced convection and active Natural cooling Fuel convection Temperatures Building Exhaust Flow Fuel Temperatures Cubicle Temperatures Cubicle to Building Flow

• Spent fuel storage tank is robust to loss of power

• Natural convection (sourced from the environment) is established and maintains adequate cooling of the fuel and storage cubicle 95

Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Loss of Power w/ Tank+Cubicle Breach Fuel Rupture Site Forced Temperatures Flow convection and active Natural cooling convection Cubicle Temperatures Cubicle to Building Flow

• Spent fuel storage tank is robust to loss of power coincident with combined tank and cubicle rupture

• Natural convection (sourced from the environment) is established and maintains adequate cooling of the fuel and storage cubicle

• Cubicle Temperatures Flow through the cubicle rupture heats building volumes 96

Spent Fuel Storage Tank Sensitivities Spent Fuel Storage Sensitivity Specification Model Parameter Distribution Range TRISO Model Fuel Pebble Emissivity (-) Uniform 0.5 - 0.999 Parameters Fuel Pebble Bed Porosity (-) Uniform 0.3 - 0.5 Graphite Conductivity Multiplier (-) Uniform 0.5 - 1.5 Decay Heat Multiplier (-) Uniform 1.0 - 10.0 Design Parameters Cubicle Flow Area Multiplier (-) Log-Uniform 0.01 - 1.0 Cubicle Filter System Discrete True/False Cubicle Fan P Uniform 1.0 - 3000.0 Tank Rupture Area Multiplier (-) Uniform 0.1 - 1.0 Scenario Parameters Cubicle Rupture Area Multiplier (-) Uniform 0.1 - 1.0 98

Quantity of Interest Horsetails Peak Fuel Peak Cubicle TRISO Failure Temperatures Temperatures Released Filtered Oxidation Products Mass Radionuclide Radionuclide Mass Mass

• Fuel Temperature drives TRISO failure and radionuclide diffusion out of TRISO

• Quantities of interest represent a large spectrum of outcomes 99

Model Sensitivities to Peak Quantities of Interest Peak TRISO Failure Peak Released Peak Fuel Radionuclide Temperatures Mass Peak Cubicle Oxidation Products Mass Temperatures Peak Filtered Radionuclide Mass

• Decay heat multiplier strongly impacts quantities of interest

• Cubicle flow area multiplier also exhibits a notable impact on quantities of interest

• Impact by other sensitivity parameters on selected quantities of interest is likely present, but smaller in magnitude and so not observed 10

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

Conclusions

• Illustrated HTGR fuel cycle modeling capabilities in MELCOR to demonstrate code readiness

• Parametric sensitivity study demonstrated the impact of UF6 cylinder rupture characteristics on material transport (i.e., vapor fraction)

• Event sensitivities indicate that used fuel storage requires large mass flows to maintain cooling on loss of power which presents a challenge for filtration

• The spent fuel storage model is robust across analyzed event sensitivities

• Parametric sensitivity study indicates that decay heat and cubicle flow blockage drive peak fuel temperatures and by extension other key quantities of interest in spent fuel storage tanks during a loss of power accident with combined tank and cubicle rupture

• Demonstrated MELCOR modeling practices for a multiple systems highlighting various stages of the HTGR fuel cycle

• Model of UF6 cylinder rupture

• Model of multiple fuel storage tank operational modes and transients

• Input of detailed ORIGEN radionuclide inventory data from ORNL

• Develop MELCOR input model for exploratory analysis

• Fast-running calculations facilitate sensitivity evaluations

• Communicated an understanding of existing non-LWR fuel cycle modeling capabilities and safety 10

Closing Remarks

• Demonstration of NRCs Code Readiness for Reviewing non-LWRs

- HTGR Nuclear Fuel Cycle

• Next Steps

- Public Reports

- SFR Workshop

Backup: Lists of scenarios for the individual stages 10

E1: Enrichment - Scenarios NUREG/CR-6410 scenarios

• E1.1 - HALEU enriched UF6 cylinder overfilled and heated UF6 release with rupture of cylinder

• E1.2 - HALEU enriched UF6 cylinder dropped UF6 liquid and vapor leaks from damaged cylinder Scenarios from National Enrichment Facility (NEF) SER

• E1.3 - Seismic or other initiating event causing pipe rupture UF6 release

• E1.4 - Fire UF6 handling hall UF6 release

• E1.5 - Unintended accumulation of enriched U inadvertent nuclear criticality 10

T1: Transportation of UF6 - Scenarios Criticality:

T1.1: Water surrounding array of canisters at optimal moderator-to-fuel ratio and optimal canister criticality T1.2: Water ingress into array of canisters at optimal moderator-to-fuel ratio criticality T1.3: Water surrounding into array of canisters with simultaneous water ingress at optimal moderator-to-fuel ratio criticality T1.4: Low ambient temperatures criticality at low temperatures T1.5: Damage to container due to drop reduced container array spacing criticality T1.6: Loss of overpack due to vehicle accident reduced container array spacing criticality Release:

T1.7: Fire due to vehicle accident melt/burn/combustion of overpack (foam insulation)

T1.8: Fire due to vehicle accident combustion of melting of plugs venting of gases T1.9: Impact due to vehicle accident rupture of container release of UF6 gas 10

F1: Fabrication of TRISO Particles - Scenarios Fire Scenarios

• F1.1 Sparks HMTA (Hexamethylenetetramine) explodes

• F1.2 Sparks HMTA catches fire

• F1.3 Heat/ignition source Uranyl nitrate solution catches fire

• F1.4 Heat/ignition source TCE explosion

• F1.5 Heat/ignition source Acetylene explosion during coating process

• F1.6 Heat/ignition source Propylene explosion during coating process

• F1.7 Heat/ignition source MTS (Methyltrichlorosilane) explosion during coating process Chemical Scenarios

• F1.8 System leak Uranyl nitrate solution thermal decomposition produces toxic nitrogen oxides which escapes into unventilated room

• F1.9 System leak Uranyl nitrate solution spill

• F1.10 System leak Silicone oil spill

• F1.11 System leak TCE (Trichloroethylene) not being ventilated (thermal decomposition leads to toxic gases and vapors)

• F1.12 System leak TCE spill

• F1.13 System leak Ammonium hydroxide decomposes to nitrogen oxides in unventilated room

• F1.14 System leak Ammonium hydroxide spill

• F1.15 Water ingress MTS reaction with water

• F1.16 System leak MTS leaks in unventilated room Criticality Scenarios

• F1.17 Improper handling of uranium nitrate hexahydrate (UNH) solution criticality

• F1.18 Flooding or water ingress oxide fuel storage criticality

• F1.19 Buildup of material in ducts or process stages criticality 10

F2: Fabrication of Fuel Pebbles - Scenarios Fire Scenarios

• F2.1 Abrasion and graphite dust Fire

• F2.2 Air ingression during heat treatment Fire Chemical Scenarios

• F2.3 Water ingress corrosion of pebbles Criticality Scenarios

• F2.4 Improper storage of fuel pebbles criticality (unexpected large enrichment, addition of moderator pebbles, water ingress, water flooding storage room, etc.)

• F2.5 Improper handling of TRISO particles criticality Downstream Considerations

• Too many damaged coated particles leading to free fuel

• Mechanical failure of pebble (cracks formed in pebble)

• Graphite impurities and density 10

T2: Transportation of Fresh Fuel Pebbles - Scenarios Criticality:

T2.1: Water surrounding array of canisters at optimal moderator-to-fuel ratio and optimal canister criticality T2.2: Water ingress into array of canisters at optimal moderator-to-fuel ratio criticality T2.3: Water surrounding into array of canisters with simultaneous water ingress at optimal moderator-to-fuel ratio criticality T2.4: Ambient temperatures vary between 40°C and 38°C criticality at low temperatures T2.5: Container drop damage to container reduced container array spacing criticality T2.6: Vehicle accident damage to container with release of fuel pebbles re-arrangement of fuel pebbles from all containers on vehicle criticality Release:

T2.7: Vehicle accident fire fire of fuel pebble graphite T2.8: Vehicle accident fire extinguishing water comes into contact with graphite at high temperature graphite corrosion and development of graphite dust 10

U1: Fresh Fuel Staging and Loading - Scenarios Criticality U1.1: Water surrounding array of canisters at optimal moderator-to-fuel ratio and optimal canister criticality U1.2: Water ingress into array of canisters at optimal moderator-to-fuel ratio criticality U1.3: Water surrounding into array of canisters with simultaneous water ingress at optimal moderator-to-fuel ratio criticality U1.4: Misplacement of array of graphite pebble and fuel pebble containers additional moderation due to graphite moderator criticality U1.5: Damage to container due to drop of container reduced container array spacing criticality U1.6: Fire in pebble handling chamber fire of fuel pebble graphite U1.7: Fire in pebble handling chamber extinguishing water comes into contact with graphite at high temperature graphite corrosion and development of graphite dust U1.8: Drop of pebbles while filling them into hopper damage of pebbles generation of graphite dust 11

U2: Power Production Including Online Refueling -

Scenarios Release:

U2.6: FHSS pipe rupture Pebbles come out out of the reactor with high temperature and pressure oxidation of graphite in contact with air pebble damage with fission product release U2.6: Fps escaped from pebbles adsorb into graphite dust (dust generated by pebble wear, fracture, irradiation sputtering, and corrosion) graphite dust flows in the primary circuit with the helium, deposits on the surface of the reactor components loss of coolant causes release of dust-gas mixture, and therefore fission product release U2.7: Air ingress into core U2.8: Chemical attack of TRISO layers and graphite (by steam) graphite oxidation U2.9: Graphite dust catches fire from sparks or heat U2.10: Broken pebble gets stuck in reactor fission product product release into He coolant 11

U2: Power Production Including Online Refueling - Scenarios Criticality:

U2.1: Failure in FHSS system additional pebbles enter core criticality U2.2: Failure in BUMS pebbles with low burnup replaced by fresh pebbles too many fresh fuel pebbles enter the core criticality U2.3: Failure in CUD pebbles are not removed from reactor, but still added on top criticality U2.4: Seismic events reorientation of pebbles (consider pebble cone in upper core) criticality U2.5: Water steam ingress into core w/o CR insertion criticality Heat removal:

• U2.11: Accumulation of hot pebbles in FHSS pipes at high temperatures and pressure (pebble jam) due to error in FHSS or stuck pebbles due to a damaged or swollen pebble temperature increase

• U2.12: depressurized loss of forced circulation (covered in Vol.3)

• U2.13: Blockage of fuel element coolant channel due graphite failure/spalling (channel distortion) temperature increase fuel pebble failure 11

U4: Onsite Discharged Pebble Storage - Scenarios Criticality:

U4.1: Graphite pebbles are misloaded into fuel pebble storage criticality U4.2: BUMS malfunction pebbles with lower burnup than discharge burnup are misloaded into fuel pebble storage criticality U4.3: Water ingress into used fuel tank criticality U4.4: Tank rupture with no tube collapse reorientation of pebbles criticality U4.5: Tank rupture with central tube collapse reorientation of pebbles criticality Heat removal:

U4.6: BUMS malfunction pebbles with higher burnup than discharge burnup are misloaded into fuel pebble storage increased temperature from decay heat U4.7: Failure of the active cooling system passive cooling system takes over through natural convection slightly higher fuel and structure temperatures U4.8: Failure of the passive cooling system because of blockage of the natural convection paths high temperature increase of fuel and structure U4.9: Dropping of pebbles within the FHSS damage of fuel pebbles pebble jammed insufficient cooling 11

U4: Onsite Discharged Pebble Storage - Scenarios Release:

• U4.10: Manufacturing defects of fuel pebbles release of fission products from defective pebbles

• U4.11: Dropping of pebbles within the FHSS damage of fuel pebbles fission product release and graphite dust

• U4.12: Dropping of pebbles inside the storage tank damage of fuel pebbles fission product release and graphite dust

• U4.13: Tank rupture with no tube collapse damage of fuel pebbles fission product release and graphite dust

• U4.14: Tank rupture with central tube collapse damage of fuel pebbles fission product release and graphite dust

• U4.15: Gamma radiation from fuel pebbles cause radiolysis of the air resulting in extremely corrosive elements such as nitric acid and ozone in the air graphite corrosion fuel pebble failure fission product release

• U4.16: Sparks from machinery, equipment, electrical circuits, or human activities fire

• U4.17: Radiolysis of the coolant air evolution of explosive gas mixtures explosion

• U4.18: Off-gassing or volatilization evolution of explosive gas mixtures explosion

• U4.19: Collision of vehicles or suspended loads with FHSS pipes pipe rupture pebble drop fission product release and graphite dust

• U4.22: Collision of vehicles or suspended loads with storage tank damage to tank damage to pebbles inside tank fission product release and graphite dust 11

Accidents Selected for Initial SCALE/MELCOR Calculations Fuel Cycle Accident SCALE/MELCOR Stage E1 - Uranium Rupture of a HALEU enriched UF6 cylinder on MELCOR - transport of UF6 Enrichment storage dock T1 - Transportation of Water ingress into array of canisters at optimal SCALE - criticality Front-end UF6 moderator to fuel ratio criticality T2 - Transportation of Damage to container due to drop of container SCALE - criticality Fresh fuel Pebbles reduced container array spacing criticality U2 - Uranium FHSS pipe rupture, pebbles exit out of the reactor Fuel SCALE - pebble inventory Utilization / Online with high temperature and pressure, leading to Handling MELCOR - release paths Refueling graphite & air interaction Collision of vehicle or suspended load with U4 - Onsite storage tank causing damage to tank and damage SCALE - spent fuel tank inventory Back-end Discharged Pebble to pebbles inside tank, causing fission product MELCOR - release paths Storage and graphite dust release 11

SCALE Backup 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

Most limiting hypothetical condition is without PSP Single cylinder (no PSP) with varying water density DN30-10 DN30-20 SCALE model of single BN30-10 package surrounded by 30 cm of water Full water density results in the most reactive configuration.

117

Reactivity sensitivity study for various VP-55 materials

• Some of the insulation material variables were varied within the specified limits to understand their impact on the criticality.

• Calculations were performed using ENDF/B-VII.1 252g MG Case Reference keff +/- sigma k (pcm)

Reference - 0.30416 +/- 0.0001 (ref) polyurethane Fiberglass type E Fiberglass type R 0.25083 +/- 0.0001 -5333 fiberglass Fiberglass type C Fiberglass type R 0.25597 +/- 0.0001 -4819 Fiberglass type R, wt = 50% 0.30078 +/- 0.0001 -338 wt=36%

Polyurethane density 11 5 PCF 0.30308 +/- 0.0001 -108 PCF*

Polyurethane with TSL h-poly** 0.29958 +/- 0.0001 -458 data of H in water**

Polyurethane without h-poly** 0.29651 +/- 0.0001 -766 TSL data of H**

• PCF: pounds per cubic foot, a unit used for measuring foam densities

• • In the absence of thermal scattering law (TSL) data for H in polyurethane, the TSL data for H in polyethylene was applied. To assess the impact on the choice of TSL data, tests were run with (1) TSL data for H in water, and (2) no TSL data for H (H as free gas).

3D SCALE VP-55 model 118

Additional water/flooding scenarios for VP-55 a) Impact of water surrounding the containers c) Impact of water ingress into the container keff much higher than air-only case b) Impact of water surrounding the containers and d) Impact of flooding: water surrounding containers inside outer drum and water ingress into the container For all cases, largest keff found when cylinders are touching (unlike the air-only case)

Impact of packing fraction on criticality for VP-55 PF = 0.45 (298 pebbles) PF = 0.55 (364 pebbles)

Optimal pitch in air is not when cylinders touching each other PF = 0.50 (331 pebbles) PF = 0.60 (397 pebbles) keff for all arrangements is far below 0.95

SCALE Criticality Calculations of VP-55 Impact of varying the enrichment on keff keff increases linearly with increasing the enrichment.

keff is more sensitive to the pitch with higher enrichment.

Decay heat of spent fuel inventory Total decay heat of pebbles with different discharge time Day of discharge 650 days after discharge 1284 days after discharge Top of tank Middle of tank Bottom of tank

Equilibrium Core Inventory Equilibrium Core Inventory Search:

1. Based on the benchmark specification in IAEA-TECDOC-1694*. Core is divided into 5 radial channels and 22 axial regions.

2. At full power and with the 24 control rods inserted 2.285 m below the bottom of the top reflector.

3. Pebble is circulated six times through the core before it is discharged.

4. After each pass, the fuel is reintroduced to the top of the core and equally distributed over any defined flow lines (or core positions).

5. Fuel flow lines are all parallel and all fuel flow speeds are the same (no variation in core residence time), independent of the radial and azimuthal position.

6. Pebble after discharged is cooled down for 7 days before re-inserted to the core.

• INTERNATIONAL ATOMIC ENERGY AGENCY, Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility, IAEA-TECDOC-1694, IAEA, Vienna (2013) 123

Rapid inventory generation of retired pebbles

• Procedures followed for generating pebble discharge inventory

1. Generate ORIGEN reactor libraries

• TRITON models were developed to generate ORIGEN reactor libraries.

• Models have information about different channels.

• Three fuel/reflector temperatures.

• Up to 100 GWd/MTU with 28 burnup steps.

2. 20,000 random pebbles histories were generated, considering different radial channel and associated power distributions

• Each history completes seven passes, each pass history is determined stochastically.

• Channel at each pass was selected based on a discrete probability distribution that accounts for the difference of the volume fraction.

3. ORIGAMI used to simulate the 20,000 histories

• 4.5 days of cooling time after the end of each pass.

• Based on the fuel/reflector temperature of each axial zone of pass, ORIGAMI calls ORIGEN libraries to interpolate problem-dependent cross-sections.

• D. Hartanto, W. A. Wieselquist , S. E. Skutnik, P. W. Gibbs (2022),"Uncertainty Quantification of Pebble's Discharge Burnup and Isotopic Inventory Using SCALE," Proc. ANS Winter Meeting, Phoenix, AZ, November 13-17, 2022.

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Decay heat of spent fuel tank Decay heat of Decay heat of spent fuel tank Total decay heat of SFT (sum of top slice that has just been fully filled all slices; all 620,000 pebbles) 125

Top contributors to decay heat of spent fuel tank Day of discharge 650 days after discharge 1284 days after discharge T1/2= 2.56 min T1/2= 64 hr T1/2= 17 min T1/2= 2.6 year T1/2= 2.56 min T1/2= 64 day T1/2= 2.4 day T1/2= 30 sec T1/2= 30 sec T1/2= 35 day T1/2= 1.7 day T1/2= 2.6 year T1/2= 17 min T1/2= 17 min T1/2= 64 hr Top of tank (3.7 kW at t=0) Middle of tank (0.06 kW at t=0) Bottom of tank (0.03 kW at t=0) 126

Spent fuel tank inventory calculation 1-Prepare discharge 2-Compute Inventory 3- Progress the 4- Generate inventory file for inventory accident for 10 days MELCOR of each slice ORIGEN ORIGEN ORIGEN OBIWAN Blend the Decay the Decay resulting inventory of discharge inventory at each Convert the binary 20,000 discharged inventory using for time step for concentration file pebbles 1284 days additional 10 days to II.JSON format

• The new ORIGEN blend block is used.

• Time step=1 day, each day

• t=0 is the immediately after the

• Blended 9/106 of each pebbles mass to represents a slice in the tank accident compute average discharge inventory of

• t=0 day last discharged (top of one pebble. the tank)

• Discharge cutoff = 85 GWd/MTU.

• t =1284 day First discharged (bottom of the tank.

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MELCOR Backup 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

HTGR Components

• Pebble Bed Reactor Fuel/Matrix Legend Components Unfueled pebble core Fueled pebble core Fueled part of pebble TRISO Unfueled shell (matrix) is modeled as Fuel (FU) separate component GRAPHITE Fuel radial temperature profile for sphere GRAPHITE Matrix (MX)

Fluid B/C

• TRISO Radionuclide Release Model Recent failures - particles failing within latest time-step (burst release, diffusion release in time-step) TRISO (FU)

Previous failures - particles failing on a previous time-step (time history of Sub-component model diffusion release) for zonal diffusion of radionuclides through Contamination and recoil TRISO particle 129

Transient/Accident Solution Methodology Stage 0: Stage 1: Stage 2: Stage 3:

Normal Operation Normal Operation Normal Operation Accident Establish thermal state Diffusion Calculation Transport Calculation Diffusion & Transport calculation Calculate steady state distribution of Calculate accident Establish steady state Establish steady state radionuclides and graphite dust progression and radionuclide temperatures and distribution of throughout system (deposition on release pressures throughout the radionuclides in TRISO surfaces, convection through flow problem domain particles and matrix paths)

Elevation [m]

Temperature [K]

Representative reflector temperature response Example:

PBMR-400 Cs 2000 K Distribution in Primary System Time [min]

Temperature [K]

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HTGR Radionuclide Diffusion Release Model Intact TRISO Particles

• One-dimensional finite volume diffusion equation solver for multiple zones (materials) Diffusivity Data Availability

• Temperature-dependent diffusion coefficients (Arrhenius form)

Porous Matrix TRISO Radionuclide UO2 UCO PyC SiC 1 Carbon Graphite Overall

=

+ = 0 Ag Some Some Extensive Some Extensive Not investigated Cs Some Some Extensive Some Some Not found I Some Some Some Not found Not found Kr Some Some Not found Some Some Sr Some Some Extensive Some Some Xe Some Some Some Some Not found Data used in the demo calculation

[IAEA TECDOC-0978]

FP Species Kr Cs Sr Ag D (m2/s) Q D (m2/s) Q D (m2/s) Q D (m2/s) Q Intact TRISO Layer (J/mole) (J/mole) (J/mole) (J/mole)

Kernel 1.3E-12 126000.0 5.6-8 209000.0 2.2E-3 488000.0 6.75E-9 165000.0 (normal)

Buffer 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 Concentrations PyC 2.9E-8 291000.0 6.3E-8 222000.0 2.3E-6 197000.0 5.3E-9 154000.0 SiC 3.7E+1 657000.0 7.2E-14 125000.0 1.25E-9 205000.0 3.6E-9 215000.0 Matrix Carbon 6.0E-6 0.0 3.6E-4 189000.0 1.0E-2 303000.0 1.6E00 258000.0 Str. Carbon 6.0E-6 0.0 1.7E-6 149000.0 1.7E-2 268000.0 1.6E00 258000.0 Iodine assumed to behave like Kr CORSOR-Booth LWR scaling used to estimate other radionuclides 131

Graphite Oxidation Steam oxidation Reactions H2O or Air He Reactions Air oxidation Both steam and air include rate limit due to steam/air diffusion towards active oxidation surface ROX is the rate term in the parabolic oxidation equation [1/s]

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COR Intercell Conduction Effective conductivity prescription for Dp=.06 m pebble bed (bed conductance) Kf=.154 W/m-K Ks = 26 W/m-K

• Zehner-Schlunder-Bauer with Breitbach-Barthels modification to the radiation term Effective conductivity prescription for prismatic (continuous solid with pores)

• Tanaka and Chisaka expression for effective radial conductivity (of a single PMR hex block)

Ks = 26 W/m-K Kf=.154 W/m-K

• A radiation term is incorporated in parallel with the pore conductivity

• Thermal resistance of helium gaps between hex block fuel elements is added in parallel via a gap conductance term 133

Interface Between Thermal-hydraulics and Pebble Bed Reactor Core Structures Heat transfer coefficient (Nusselt number) correlations for pebble bed convection:

• Isolated, spherical particles

• Use Tfilm to evaluate non-dimensional numbers, use maximum of forced and free Nu 4 3 3

= 2.0 + 0.6 1 1 = 2.0 + 0.6 1 2 1

• Constants and exponents accessible by sensitivity coefficient Flow resistance

• Packed bed pressure drop Loss coefficient relative to Ergun (original) coefficient at Re=1000 1 4

, = 1 + 2 1

+ 3 1

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