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

• 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 Outline

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

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 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

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

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

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

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 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

11 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 Overview

12 HTGR Fuel Cycle

- Enrichment

- Gas centrifuges

- Transportation of UF6

- Package: DN30-X

- TRISO fabrication

- Sol-gel process

- Pebble fabrication

- Process: X-energy

- Transportation of fuel pebbles

- Package: Versa-Pac

- Fresh fuel staging

- Pebble loading

- Power production

- Online refueling

- Discharged pebble storage onsite

- Spent fuel/used fuel/graphite tanks E1 T1 F1 F2 T2 U1 U2 U4 Not covered: uranium mining & milling, spent fuel transportation & off-site storage & final disposal

13 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 E1: Enrichment Major hazards:

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

14 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):

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 ORANO: 30B cylinder with DN30 PSP Ref.: ORANO Safety Analysis Report for the DN30-X Package https://www.nrc.gov/docs/ML2232/ML22327A183.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

15 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:

Coat the kernels with the carbon layers using various gas mixtures at different temperatures 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).

Ref.: P. Pappano, TRISO-X Fuel Fabrication Facility Overview, Introductory Meeting with the NRC, ML18254A086 (2018). https://www.nrc.gov/docs/ML1825/ML18254A086.pdf 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

16 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 Ref 1: IAEA, Fuel performance and fission product behavior, IAEA-TECDOC-978 (1997).

Ref 2: "PBMR Coupled Neutronics / Thermal-hydraulics Transient Benchmark The PBMR-400 Core Design, Vol. 1: The Benchmark Definition," NEA/NSC/DOC(2013)10, 2013.

Major hazards:

Criticality due to improper storage of TRISOs or fuel pebbles Contact with water leading to graphite corrosion Development of graphite dust leading to fire hazard PBMR-400 fuel pebble and TRISO particle2

17 T2: Transportation of Fresh Fuel Pebbles Versa-Pac Ref.: DAHER-TLI Versa-Pac Safety Analysis Report https://www.nrc.gov/docs/ML1833/ML18330A093.pdf 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:

Criticality due to water accidents and container drop Contact with water leading to graphite corrosion Development of graphite dust leading to fire hazard

18

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)

U1/U2/U4 - Utilization Stages 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.

19 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 U1: Fresh Fuel Staging and Loading Major hazards:

Criticality due to water accidents, graphite pebble misloading, tank rupture Development of graphite dust leading to fire hazard

20 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 U2: Power Production Including Online Refueling Major hazards:

Criticality due to pebble misloading, incorrect burnup measurement, failed core unloading device Temperature increase in pipes or core due to stuck pebbles Fission product release into coolant or adsorption into graphite dust Graphite oxidation due to chemical attack Ref.: C. C. Stoker et al. PBMR Used Fuel Storage Criticality for Most Reactive Core Loading. Proc. ICNC, St. Petersburg, Russia, 28 May-1June, pages 8-14, 2007.

21 10 Spent Fuel Tank (SFT):

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.

U4: Onsite Discharged Pebble Storage Ref.: C. C. Stoker et al. PBMR Used Fuel Storage Criticality for Most Reactive Core Loading. Proc. ICNC, St. Petersburg, Russia, 28 May-1June, pages 8-14, 2007.

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 Development of graphite dust leading to fire hazard Ref.: J. Slabber. Reactor Unit and Main Support Systems.

https://www.nrc.gov/docs/ML0 606/ML060680079.pdf, 2006

22 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 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 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

24 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 Objective and applications

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

25 Reference HTGR: PBMR-400 PBMR-400 SCALE model [2]

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

[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)

PBMR-400 TRISO particle and fuel pebble [1]

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 Scenario 1: Water ingress into packages during UF6 transportation

27 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) 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 DN30-X UF6 transportation package DN30-X package*

30B-X cylinder*

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

28 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)

SCALE model of DN30-X package 3D view of SCALE model 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 libraries

• The same assumptions in the safety analysis report are adopted: Safety Analysis Report for the DN30-X Package, https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML22327A183 UF6 PSP Control rod

29 SCALE baseline result for DN30-X 3D view of SCALE model*

• The same assumptions in the safety analysis report are adopted: Safety Analysis Report for the DN30-X Package, https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML22327A183 UF6 PSP Control rod 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 30B-10 (33 control rods) 30B-20 (43 control rods)

Infinite hexagonal array of packages touching on sides, surrounded by airno water ingress.

30 Impact of water on criticality for DN30-10 Water surrounding PSP Water surrounding and inside PSP Minimum package pitch (touching) is the most reactive configuration.

Water ingress into PSP has lower keff.

31 Impact of water on criticality for DN30-20 Water surrounding PSP Water surrounding and inside PSP DN30-20 shows the same trends as DN30-10.

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

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 Scenario 2: Damage/drop during fresh fuel pebble transportation

33 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 Container permits maximum of 584 pebbles based on given enrichment, and up to 505 g of 235U permitted loading. 364 pebbles fit into container at 55% packing fraction.

Versa-Pac Illustration*

• Versa-Pac Safety Analysis Report, https://www.nrc.gov/docs/ML1833/ML18330A093.pdf Fresh pebble transportation package Characteristic Value Fuel enrichment 9.6 wt.% 235U TRISO packing fraction

~9%

Uranium content per pebble 9 g

34 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 recommended limits SCALE model of the VP-55 3D SCALE VP-55 model

35 SCALE baseline result for the VP-55 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

Reference:

Infinite array of touching containers surrounded by air

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

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 Scenario 3: Pebble ejection from fuel handling system

39 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 SCALE approach for fuel inventory generation

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

[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.

SCALE PBMR-400 model

Characteristics of pebbles in PBMR-400 Average pebble burnup after each pass Burnup distribution after each pass The error bars correspond to the burnup range after each pass Target burnup is 90 GWd/MTU, but 7 passes are simulated to include pebbles that havent reached the target burnup at 6th pass.

Pebble burnup [GWd/MTU]

Target burnup and number of passes Fraction of retired pebbles using cut-off value 85 GWd/MTU On average, it takes 6 passes through the core for a pebble to reach the target burnup of 90 GWd/tU.

With burnup limit at BUMS of 85 GWd/MTU, the average burnup of the retired pebbles is 90 GWd/MTU (target burnup).

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

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

43 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

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 Scenario 4: Collision with the spent fuel storage tank

45

• 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 spent fuel 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

46

• 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 SFT modeling procedure Spent fuel tank

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

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

The top layers are dominating decay heat and the sharp drop is driven by nuclides in that range.

Top Layer Decay Heat Top Layer Contributors

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 Summary

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

51

• 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.

Conclusions of SCALE analysis

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

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

53 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 Corrosives Other phenomenological models Graphite oxidation Intercell and intracell conduction Convection & flow Control function-based chemistry MELCOR HTGR Fuel Cycle Modeling Fuls, W.F., Mathews, E.H. (2006). Passive Cooling of the PBMR spent and used fuel tanks. Nuclear Engineering and Design, 237, 1354-1362.

Slabber (2006), Technical Description of the PBMR Demonstration Power Plant, https://www.nrc.gov/docs/ML0609/ML060940293.pdf

54 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 Capability Demonstration

MELCOR Models and Simulations

56 A Short Summary of Facility Modeling with MELCOR Images of the title pages for -

barnwell nsrd-10 Source term and leak path factor analysis (aerosol physics, vapor physics, user-defined speciation and chemistry, etc)

Broad accident sequence spectrum (multi-room fire, explosions, spills, etc.)

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

MELCOR capabilities facilitate radiological and non-radiological hazard analyses

Facility

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

Power conversion compartment (PCC)

Power conversion crane compartment (PCCC)

Reactor compartment (RC)

Reactor crane compartment (RCC)

Storage compartment (SC)

Auxiliary Compartment (AC)

59 Facility Model Detail Storage Compartment Auxiliary Compartment Stairwell Exhaust Intake Environment 90 61

-10

-2 30 0

4 Altitude [m] (not to scale) 26 23 Building Filter Supply Flow Exhaust Flow Doors

• Flow connections not representative of connection altitudes Compartment Volume [m3]

Environment 1000.0 Intake 1000.0 Stairwell 3200.0 Auxiliary Compartment 12400.0 Storage Compartment 38400.0 Building Filter 10.0 Exhaust 1000.0

60 Building Filter Detail 30 23 Altitude [m] (not to scale) 26 Pre-filter HEPA Filter Fan Inlet Fan Outlet To Building Exhaust From Building Compartments Supply Flow Exhaust Flow Fan

• Flow connections not representative of connection altitudes Compartment Volume [m3]

Pre-filter 5.0 HEPA 2.0 Fan Inlet 2.0 Fan Outlet 2.5 Specification Fan P [Pa]

100.0

UF6 Cylinder

62 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 E1: Enrichment - UF6 Cylinder Rupture

63 Demonstration UF6 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

UF6 Cylinder Detail 90 61

-10

-2 30 0

4 Altitude [m] (not to scale) 26 23 UF6 Storage Supply Flow Exhaust Flow Doors

• Flow connections not representative of connection altitudes Compartment Volume [m3]

UF6 Storage 1000.0 Specification UF6 release mass

[kg]

14000 Flashing Ratio 0.45 vapor/0.55 aerosol Building Relative Humidity 0.4 Release Duration [s]

1.0 x 10-3 Door Open Fraction 1.0

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

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

UF6 Cylinder Sensitivities

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

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

70 Fuel Storage Tank

71 Demonstration Fuel Storage Inlet Exhaust Concrete Cubicle Gamma Shield Thermal Shield Downcomer Storage Tank Cooling Tubes Fuel Spheres Fuls, W.F., Mathews, E.H. (2006). Passive Cooling of the PBMR spent and used fuel tanks. Nuclear Engineering and Design, 237, 1354-1362.

MELCOR fuel storage tank concept overview with designated coolant flow Spent Fuel: Retired fuel that has reached a specified burnup and cannot be reloaded into the core Used Fuel: Fuel that has not reached the specified burnup and can be reloaded into the core May require temporary storage during core maintenance

72 Demonstration Fuel Storage: Operational Modes Closed Loop Active Cooling Normal operational mode for spent fuel storage tanks nominal decay heat ~140kW Building flow is isolated from concrete cubicle flow Open Loop Active Cooling Normal operational mode for used fuel storage tanks nominal decay heat ~640kW Concrete cubicle draws on building air supply Open Loop Passive Cooling On loss of power, louvres open (transition from closed to open loop) and/or active cooling is lost (spent or used fuel, respectively)

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

73 Cubicle Model Additions 90 61

-10

-2 30 0

4 Altitude [m] (not to scale) 26 23 Cubicle Filter Fuel Storage Cubicle Supply Flow Exhaust Flow Doors Rupture

• Flow connections not representative of connection altitudes

74 Reference Diagram with design flow Fuel Storage Cubicle Detail Closed Loop Flow Open Loop Flow Rupture

• Flow connections not representative of connection altitudes Top Storage Tank Bypass 90 61

-2 30 0

4 Altitude [m] (not to scale) 26 23 Cubicle Inlet Cubicle Exhaust Plenum Bottom Lower Middle Downcomer Storage Compartment Cubicle Filter

75 Fuel storage tank radial nodalization Fuel storage tank axial nodalization Fuls, W.F., Viljoen, C., Stoker, C., Koch, C., Kleingeld, M. (2005). The interim fuel storage facility of the PBMR.

Annals of Nuclear Energy, 32, 1854-1866.

Fuel Storage Tank Detail 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 Difference

76 Cubicle Filter Detail 30 23 Altitude [m] (not to scale) 26 Pre-filter HEPA Filter Fan Inlet Fan Outlet To Building Exhaust From Cubicle Exhaust Supply Flow Exhaust Flow Fan

• Flow connections not representative of connection altitudes Compartment Volume [m3]

Pre-filter 5.0 HEPA 2.0 Carbon 2.0 Fan Inlet 2.0 Fan Outlet 2.5 Specification Fan P [Pa]

100.0 Carbon Filter

77 Postulated Scenario Event Tree MELCOR flexibility facilitates exploration of large event spaces

78 Used Fuel Storage Tank

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

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

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

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

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

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

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

86 Spent Fuel Storage Tank

87 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 U4: Discharged Pebble Storage - Spent Fuel Storage Tank Transients

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

89 Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Tank Rupture Rupture Site Flow Forced convection and active cooling Spent fuel storage tank is robust to a tank breach Adequate cooling of the fuel and storage cubicle is maintained Cubicle Flow Cubicle Temperatures Fuel Temperatures

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

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

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

93 Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Heat Exchanger Failure Cubicle Cooling Forced convection without heat removal Spent fuel storage tank is challenged by loss of active cooling Without active cooling, the fuel and cubicle atmosphere heats up Cubicle Flow Cubicle Temperatures Fuel Temperatures

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

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

96 Fuel Storage Tank - Spent Fuel w/ Active Closed Loop Heat Removal w/ Loss of Power w/ Tank+Cubicle Breach Cubicle Temperatures 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 Flow through the cubicle rupture heats building volumes Natural convection Forced convection and active cooling Cubicle Temperatures Fuel Temperatures Rupture Site Flow Cubicle to Building Flow

Spent Fuel Storage Tank Sensitivities

98 Spent Fuel Storage Sensitivity Specification Model Parameter Distribution Range TRISO Model Parameters Fuel Pebble Emissivity (-)

Uniform 0.5 - 0.999 Fuel Pebble Bed Porosity (-)

Uniform 0.3 - 0.5 Design Parameters Graphite Conductivity Multiplier (-)

Uniform 0.5 - 1.5 Decay Heat Multiplier (-)

Uniform 1.0 - 10.0 Cubicle Flow Area Multiplier (-)

Log-Uniform 0.01 - 1.0 Cubicle Filter System Discrete True/False Cubicle Fan P Uniform 1.0 - 3000.0 Scenario Parameters Tank Rupture Area Multiplier (-)

Uniform 0.1 - 1.0 Cubicle Rupture Area Multiplier (-)

Uniform 0.1 - 1.0

99 Quantity of Interest Horsetails Fuel Temperature drives TRISO failure and radionuclide diffusion out of TRISO Quantities of interest represent a large spectrum of outcomes Filtered Radionuclide Mass Peak Fuel Temperatures Peak Cubicle Temperatures Released Radionuclide Mass Oxidation Products Mass TRISO Failure

10 0

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

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

10 2

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 Conclusions

Closing Remarks

• Demonstration of NRCs Code Readiness for Reviewing non-LWRs

- HTGR Nuclear Fuel Cycle

• Next Steps

- Public Reports

- SFR Workshop

10 4

Backup: Lists of scenarios for the individual stages

10 5

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 6

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 T1: Transportation of UF6 - Scenarios

10 7

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 F1: Fabrication of TRISO Particles - Scenarios

10 8

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 9

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 T2: Transportation of Fresh Fuel Pebbles - Scenarios

11 0

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 1

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 2

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 U2: Power Production Including Online Refueling - Scenarios

11 3

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 4

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 U4: Onsite Discharged Pebble Storage - Scenarios

11 5

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

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 Backup

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

118 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 Reactivity sensitivity study for various VP-55 materials Case Reference keff +/- sigma k (pcm)

Reference 0.30416 +/- 0.0001 (ref)

Fiberglass type E Fiberglass type R 0.25083 +/- 0.0001

-5333 Fiberglass type C Fiberglass type R 0.25597 +/- 0.0001

-4819 Fiberglass type R, wt=36%

wt = 50%

0.30078 +/- 0.0001

-338 Polyurethane density 11 PCF*

5 PCF 0.30308 +/- 0.0001

-108 Polyurethane with TSL data of H in water**

h-poly**

0.29958 +/- 0.0001

-458 Polyurethane without TSL data of H**

h-poly**

0.29651 +/- 0.0001

-766

• 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).

fiberglass polyurethane 3D SCALE VP-55 model

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

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

PF = 0.50 (331 pebbles)

PF = 0.55 (364 pebbles)

PF = 0.60 (397 pebbles)

Optimal pitch in air is not when cylinders touching each other keff for all arrangements is far below 0.95

SCALE Criticality Calculations of VP-55 Impact of varying the enrichment on keff keff is more sensitive to the pitch with higher enrichment.

keff increases linearly with increasing the enrichment.

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

123 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.

Equilibrium Core Inventory

• 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)

124 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.

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

Decay heat of top slice

126 Top contributors to decay heat of spent fuel tank 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)

Day of discharge 650 days after discharge 1284 days after discharge T1/2= 17 min T1/2= 1.7 day T1/2= 35 day T1/2= 64 day T1/2= 17 min T1/2= 30 sec T1/2= 2.6 year T1/2= 64 hr T1/2= 2.56 min T1/2= 2.56 min T1/2= 17 min T1/2= 30 sec T1/2= 2.6 year T1/2= 64 hr T1/2= 2.4 day

127 Spent fuel tank inventory calculation Blend the inventory of 20,000 discharged pebbles Decay the discharge inventory using for 1284 days Decay resulting inventory at each time step for additional 10 days Convert the binary concentration file to II.JSON format The new ORIGEN blend block is used.

Blended 9/106 of each pebbles mass to compute average discharge inventory of one pebble.

Discharge cutoff = 85 GWd/MTU.

Time step=1 day, each day represents a slice in the tank t=0 day last discharged (top of the tank) t =1284 day First discharged (bottom of the tank.

t=0 is the immediately after the accident 1-Prepare discharge inventory 2-Compute Inventory of each slice 3-Progress the accident for 10 days ORIGEN ORIGEN ORIGEN OBIWAN 4-Generate inventory file for MELCOR

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 MELCOR Backup

129

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

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

Previous failures - particles failing on a previous time-step (time history of diffusion release)

Contamination and recoil HTGR Components Legend TRISO (FU)

Fuel (FU)

Matrix (MX)

Fluid B/C TRISO GRAPHITE Sub-component model for zonal diffusion of radionuclides through TRISO particle GRAPHITE Fueled pebble core Unfueled pebble core

130 Transient/Accident Solution Methodology Stage 1:

Normal Operation Diffusion Calculation Establish steady state distribution of radionuclides in TRISO particles and matrix Stage 2:

Normal Operation Transport Calculation Calculate steady state distribution of radionuclides and graphite dust throughout system (deposition on surfaces, convection through flow paths)

Example:

PBMR-400 Cs Distribution in Primary System Stage 3:

Accident Diffusion & Transport calculation Calculate accident progression and radionuclide release Elevation [m]

Temperature [K]

2000 K Stage 0:

Normal Operation Establish thermal state Establish steady state temperatures and pressures throughout the problem domain Temperature [K]

Time [min]

Representative reflector temperature response

131 Intact TRISO Particles

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

• Temperature-dependent diffusion coefficients (Arrhenius form)

HTGR Radionuclide Diffusion Release Model Intact TRISO Concentrations

= 1

+

Layer FP Species Kr Cs Sr Ag D (m2/s)

Q (J/mole)

D (m2/s)

Q (J/mole)

D (m2/s)

Q (J/mole)

D (m2/s)

Q (J/mole)

Kernel (normal) 1.3E-12 126000.0 5.6-8 209000.0 2.2E-3 488000.0 6.75E-9 165000.0 Buffer 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 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 Data used in the demo calculation

[IAEA TECDOC-0978]

= 0

Diffusivity Data Availability Radionuclide UO2 UCO PyC Porous Carbon SiC Matrix Graphite TRISO Overall Ag Some Not investigated Some Not found Extensive Some Extensive Cs Some Some Extensive Some Some 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 Iodine assumed to behave like Kr CORSOR-Booth LWR scaling used to estimate other radionuclides

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

133 Effective conductivity prescription for pebble bed (bed conductance)

COR Intercell Conduction

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

• 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 Effective conductivity prescription for prismatic (continuous solid with pores)

• Zehner-Schlunder-Bauer with Breitbach-Barthels modification to the radiation term Dp=.06 m Kf=.154 W/m-K Ks = 26 W/m-K Ks = 26 W/m-K Kf=.154 W/m-K

134 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

• Constants and exponents accessible by sensitivity coefficient Interface Between Thermal-hydraulics and Pebble Bed Reactor Core Structures Flow resistance

• Packed bed pressure drop

, = 1 + 2 1

+ 3 1

4 1

Loss coefficient relative to Ergun (original) coefficient at Re=1000

= 2.0 + 0.6

1 4

1 3

= 2.0 + 0.6

1 2

1 3