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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Impact of trapping on tritium self-sufficiency and tritium inventories in fusion power plant fuel cycles

The dynamic analysis of fusion power plant (FPP) fuel cycles highlights the challenge of achieving tritium self-sufficiency in future FPPs. While state-of-the-art fuel cycle models offer valuable insights into the necessary design parameters for attaining tritium self-sufficiency, none of these models currently consider the impact of tritium trapping within fuel cycle components. However, detailed analysis of individual components reveals that substantial amounts of tritium can be trapped within the first wall, divertors, and breeding blanket systems, suggesting that tritium trapping may significantly influence the FPP ability to achieve self-sufficiency. The compounded effects of additional tritium traps generated by irradiation effects and component replacements further exacerbate this challenge. The novelty of this work is the integration of an explicit, physics-based model for tritium trapping, evolution of damage-induced traps, and component replacements into a dynamic, system-level model of a fuel cycle. The results show an increase of a factor 10 3 – 10 4 of tritium inventory in the first wall and vacuum vessel of an ARC-class FPP when accounting for the aforementioned phenomena. This, coupled with the replacement of components subject to significant tritium trapping, slows down fuel cycle dynamics, resulting in an extended tritium doubling time (50% increase), higher start-up inventory (30% increase), and higher required tritium breeding ratio (2%–5%) compared to a scenario without tritium trapping.

fuel cycle↗
DOE OSTI2025

Balmer emission measurements in JET-ILW hydrogen, deuterium, tritium and deuterium-tritium low-confinement mode plasmas

Measurements of Balmer-alpha (B-α) and Balmer-gamma (B-γ), where B = H, D, or T isotopes of hydrogen, emission across the low-field side (LFS) divertor in JET ITER-like Wall (ILW), low-confinement mode (L-mode) plasma indicate that the divertor plasma was more strongly detached in deuterium, tritium and deuterium–tritium than in hydrogen plasmas. These measurements are consistent with the lower degree of detachment in hydrogen versus tritium plasmas inferred from Langmuir probes (Groth et al., 2023). The ratio of B-γ to B-α is a suitable indicator of high density/low temperature conditions because B-α emission is dominated by electron impact excitation with a small electron–ion recombination component regardless of electron temperature and density (for conditions in tokamak plasmas) whilst for B-γ the recombination component increases and the excitation component decreases as the electron temperature decreases and the electron density increases. Within the uncertainty of the measurements, the electron density at the low-field side separatrix and the power crossing the separatrix are similar to permit these comparisons. The B-©/B-α ratio is well correlated with the onset of detachment as measured by Langmuir probes (McCracken et al., 1998; Groth et al., 2023). The ratios of B-γ and B-α varied between 0.005 and 0.035 for low-recycling, high-recycling and partially detached conditions, and were 60–70 % higher in partially detached hydrogen than in tritium plasmas. In contrast, the difference between deuterium, tritium and 40–60 % deuterium–tritium is within 10 % and thus within the uncertainties of the measurements. Profiles of the B-α and B-γ emission was measured with two filtered CCDs connected to the visible lost light leg of JET’s mirror link spectroscopy system (Meigs et al., 2010; Lomanowski et al., 2014). Spatially (radial) narrower B-α profiles were measured with increasing isotope mass, while similarly broadened B-γ profiles were measured.

Balmer↗
DOE OSTI2025

MOOSE-based Tritium Migration Analysis Program, Version 8 (TMAP8) for advanced open-source tritium transport and fuel cycle modeling

Tritium management is critical for the safety, sustainability, and economics of fusion energy systems, and advanced and reliable modeling tools help accelerate the development of tritium technologies. This paper presents the Tritium Migration Analysis Program, Version 8 (TMAP8), an open-source, MOOSE-based application developed to provide state-of-the-art tritium transport and fuel cycle modeling capabilities. TMAP8 aims to expand the capabilities of previous versions (i.e., TMAP4 and TMAP7) by leveraging modern computational techniques, ensuring high software quality assurance standards (key to building trust), and enabling multispecies, multiscale, and multiphysics simulations for integrated tritium transport modeling in complex geometries. This paper outlines TMAP8’s scope and rigorous development practices, emphasizing its transparency, accessibility, modularity, and reliability. We present the current suite of verification and validation cases based on those from TMAP4, demonstrating TMAP8’s accuracy and reliability against analytical solutions and experimental data. Additionally, the paper showcases TMAP8’s integrated fuel cycle modeling capabilities, highlighting its applicability at various scales and levels. The TMAP8 code and documentation are openly available, promoting collaborative development and widespread adoption within the fusion community. Future work will soon expand TMAP8’s verification and validation suite to include those from TMAP7 and other recent experimental studies for validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Implementation of tritium transport in a gas-liquid contactor CFD simulation of tritium extraction from lead-lithium in ANSYS fluent

Modifications to the Computational Fluid Dynamic (CFD) software ANSYS Fluent were done to quantify and characterize tritium transport in Gas-Liquid Contactors (GLCs). A double-slit, Ergun-like equation was employed for the porous media model, with Ergun coefficients validated with Sulzer’s Sulcol software. Tritium transport from PbLi within the GLC was verified against analytical models. The geometry of the CFD model was based on the MELODIE GLC experiment. The hydrodynamic CFD pressure drop results align well with SulCol estimations and fall between the predictions of the analytical Delft-Olujic and Billet & Schultes models. In terms of mass transfer efficiency, traditional mass transfer models showed a significant deviation from experimental results when using varying values of H solubility in PbLi. A saturation phenomenon occurred when utilizing high solubility values for hydrogen in PbLi. In conclusion, a modified film theory mass transfer coefficient, incorporating either the Delft-Olujic or Billet & Schultes wettability model, yielded CFD-predicted extraction efficiencies that closely matched experimental measurements.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

First principles density functional theory study of tritium species adsorption on Ni(111) surface and diffusion in nickel-sublayer for tritium storage

The nickel-plated zircaloy-4 is used as a tritium ( 3 H) getter in the tritium-producing burnable absorber rods (TPBARs) to capture 3 H produced in the 6 Li-riched annular γ-LiAlO 2 pellet under neutron irradiation. The experimental data and our previous theoretical results showed that the 3 H species produced from the γ-LiAlO 2 pellet were mainly 3 H 2 and 3 H 2 O. These 3 H species diffuse from the surface of the LiAlO 2 pellet across vacuum to the nickel-plated zircaloy-4 getter and then further diffuse into the getter to chemically form metal hydrides. While a number of studies show that oxygen binds strongly as compared to 3 H on the nickel (Ni) layer, the detailed mechanism of 3 H species absorption and diffusion across the Ni plate and Ni/Zr interface are still unclear. By employing density functional theory calculations, here we explored the 3 H 2 and 3 H 2 O species adsorption and dissociation on the Ni(111) surface and diffusion into the Ni sublayer. Our results indicated that the 3 H 2 and 3 H 2 O dissociate on the Ni(111) surface. The NiO x and Ni(O 3 H) x could be formed in the Ni layer due to the higher oxygen (O) diffusion energy barrier and formation of Ni vacancy defects. The oxygen was found to be retained in the Ni layer from diffusing across the Ni–Zr interface. This was revealed by comparing the diffusion barriers for 3 H with O. 3 H was found to have nearly three times smaller diffusion barrier than for O, making 3 H comparatively easier to diffuse through the Ni layer. In conclusion, the obtained results provide guidelines for experimental measurements on 3 H retention behavior in TPBARs and may open further avenues to explore the impurity effects on 3 H diffusion and storage at the Ni/zircaloy interfaces.

Tafen, De Nyago [National Energy Technology Lab. (↗
DOE OSTI2025

Evaluating the Impact of Tritium Permeation Membrane Performance and Direct Internal Recycling on Fusion Fuel Cycle Efficiency Using TMAP8

An efficient fuel cycle is vital to sustainable and cost-effective energy generation in fusion systems. Since tritium is not widely available, fusion systems must breed their own tritium for sustainable fusion deuterium-tritium reactions. An inefficient fuel cycle increases the tritium inventory needed for operations, which increases costs, constraints on tritium management systems, and safety concerns. A fuel cycle model is a powerful tool for understanding tritium inventories and flow rates across all systems in the fuel cycle. By simplifying the technical details into time-dependent tritium flow rates and inventories, the model can simulate the entire fuel cycle with high computational efficiency, even for technologies that are still under development. It can therefore quantify the impact of new tritium management technologies on fuel cycle efficiency. To evaluate the impact of key components on reducing tritium inventory, we are using and expanding an existing fuel cycle models based on latest advancements in fuel cycle research. The new model integrates Tritium Permeation Membrane (TPM) and Direct Internal Recycling (DIR) to enhance tritium transport from blanket breeders and plasma exhaust. These fuel cycle models are implemented in TMAP8 (Tritium Migration Analysis Program, version 8), a MOOSE-based open-source application designed to provide cutting-edge capabilities for tritium transport and fuel cycle modeling. The study aims to demonstrate the extensibility of existing fuel cycle modeling capability in TMAP8 and to offer a proof-of-principle design for future fusion plant systems. The presentation will cover the performance of fuel cycle modeling capabilities available in TMAP8, highlight advancements in fuel cycle research, and present a sensitivity analysis of these models. The results underline potential approaches and technology solutions to lower tritium inventory requirements, highlighting their role in shaping the future of fusion energy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effect of Mg and Ni impurities on tritium diffusion in lithium ceramics through cluster dynamics simulations

This study investigates the impact of Mg and Ni doping on tritium diffusion in LiAlO 2 and LiAl 5 O 8 ceramics, that are used in tritium-producing burnable absorber rods (TPBARs). Utilizing Centipede simulations across a broad temperature range (500 K to 1250 K), we explore the interplay between defect dynamics, cluster formation, and tritium mobility. In LiAlO 2 , Mg doping significantly enhances tritium diffusivity by increasing tritium interstitial concentrations and diffusion coefficients of key species, thereby doubling the overall tritium diffusivity. Ni doping, while shifting the dominant defect to Li vacancies, maintains high tritium mobility due to the low binding energy of Li vacancy-tritium complexes, which ensures effective tritium migration. In LiAl 5 O 8 , Mg and Ni doping results in a slight reduction in the diffusion coefficients of key species, yet the dramatic increase in tritium interstitial concentrations compensates, leading to a net small increase in tritium diffusivity. In conclusion, the findings highlight the critical role of defects in tritium transport and the effect of Mg and Ni defects on the performance of these ceramics in demanding nuclear environments.

Cluster dynamics↗
DOE OSTI2025

Logical approach to tritium vacuum pump selection for fusion applications

Vacuum pumps have been used for tritium processing since the 1940s following the isolation of pure tritium. The largest problem facing tritium vacuum pumps is the tritium radioactive decay. The beta particles, recoil energy, and subsequent radical formations are detrimental to most carbon-based materials. Carbon materials, such as plastics, elastomers, carbon fiber, activated carbon, etc., react with tritium to form methane and in the process strip material from the carbonbased material. Other reactions can be self-catalyzing, as is the case with nitrogen and tritium forming tritiated ammonia at room temperature and pressure. Tritium also readily isotopically exchanges with surface -OH groups, increasing the amount of protium in the gas stream while creating a tritium holdup on the surface. Metals are the preferred material when working with tritium, but tritium will permeate into and through almost all metals at some point. The entrapped tritium can decay inside the metal, creating local helium distortions that can impact the structural properties above and beyond the hydrogen embrittlement issue tritium can cause.

Angelette, Lucas M. [Savannah River National Labor↗
DOE OSTI2025

Tritium Breeder Composites for Fusion Applications

Due to the short half-life of tritium, all fusion reactors planned for long-term operation using a deuterium-tritium fuel cycle must produce sufficient tritium from reactions outside the core to maintain reactor operation. Typically, the core is surrounded by a tritium breeder blanket filled with lithium-containing materials which react under a neutron flux to form tritium that can then be collected. Here, reactor designers are forced to choose between using chemically reactive liquid breeders (e.g., Pb-Li, FLiBe, Li) or solid breeders, which often have lower lithium concentrations and questionable mechanical integrity under irradiation (e.g., Li4SiO4, Li2TiO3). To expand this pool of options, this work has sought to produce composite tritium breeder materials using lithium-containing ceramics and metal reinforcements. The resultant composite tritium breeders simultaneously feature higher lithium densities and improved mechanical integrity, including tensile strength and toughness, compared to other tritium breeders previously reported in literature.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Grain Boundary Effects on Radiation Damage and Tritium Diffusion in Li–Al–O Ceramics from Molecular Dynamics and Experiments

Understanding irradiation damage and tritium transport in LiAlO2 ceramics is essential for their deployment in tritium-producing burnable absorber rods (TPBARs). Grain boundaries (GBs) play an important role in governing radiation response and tritium transport in LiAlO2 and its secondary phase LiAl5O8. Using molecular dynamics simulations, we investigated defect evolution and tritium diffusion during displacement cascades in single-crystal and bicrystal LiAlO2 and LiAl5O8 at 600 K. The results reveal that GBs suppress the damage in the LiAlO2 bicrystals as compared to single crystals, via interstitial emission. The effect is more pronounced for Li defects, than for Al and O. In LiAl5O8 the damage reduction due to introduction of GB is less pronounced overall. Tritium diffusion coefficients at GBs are higher by factors of 2–10 relative to diffusion inside grain bulk. The effect is particularly pronounced in LiAlO2, where S5, S17, and S25 GBs promote rapid tritium transport, whereas LiAl5O8 exhibits slower diffusion due to reduced free volume at its GBs. These findings evince a trade-off that while GBs mitigate radiation damage by absorbing interstitials, they simultaneously provide fast pathways for tritium migration which is not desirable for TPBAR application. The mechanistic insights gained here establish a foundation for microstructural design strategies to balance radiation tolerance and tritium retention in ceramic breeder materials.

Roy, Ankit↗
DOE OSTI2025

Modelling Tritium Production and Release at High-Energy Proton Accelerators

Tritium is a well-known byproduct of particle accelerator operations. To keep levels of tritium below regulatory limits, tritium production is actively monitored and managed at Fermilab. We plan to study tritium production in the targets, beamline components, and shielding elements of the Fermilab facilities such as NuMI, BNB, and MI-65. To facilitate the analysis, we construct a simple model and use three Monte Carlo radiation codes, FLUKA, MARS, and PHITS, to estimate the amount of tritium produced in these facilities. The analysis could also serve as an intercomparison between these code results related to tritium production. To assess the actual amounts of tritium that would be released from various materials, we employ a semi-empirical diffusion model. The results of this analysis are compared to experimental data whenever possible. This approach also helps to optimize proposed target materials with respect to the tritium production and release.

Georgobiani, Dali [Fermilab]↗
DOE OSTI2025

New class of tritium breeders for fusion applications: Metal-reinforced composite breeders

Commercial fusion reactors operating on a D-T fuel cycle will require a steady supply of tritium to maintain the burning plasma required for continuous power generation. Since tritium has a short half-life, there is negligible natural abundance which necessitates fusion reactors to produce their own source of tritium. Tritium is most easily produced by surrounding a fusion reactor core with lithium (Li), which reacts under the intense neutron flux leaving the reactor core to form tritium and helium. Due to the hazards and technical challenges associated with surrounding a fusion reactor core with many tons of molten Li, other Li-bearing tritium breeder materials have been pursued. Unfortunately, most of the liquid breeders historically examined are exceedingly corrosive to reactor structural materials while many solid breeders in the form of ceramics are forced to make tradeoffs between Li content and mechanical integrity. In this work, to break the historic limit between Li-density and mechanical integrity of traditional solid breeders, a new class of solid tritium breeders is developed: metal-reinforced composite (MERC) breeders. Specifically, the high Li-density of lithium oxide (Li 2 O) is exploited through the addition of a metal reinforcing phase, producing a composite breeder material exhibiting high splitting tensile strength and quasi-ductility with a Li-density greater than other leading solid breeder candidates, including lithium orthosilicate (Li 4 SiO 4 ) and lithium metatitanate (Li 2 TiO 3 ). Mechanical testing, microstructural characterization, and neutronic simulation results are presented and discussed in light of fusion reactor design considerations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Techno-Economic Analysis of Gas-Liquid Contactors for Tritium Extraction from Lead-Lithium

To enable a sustainable fuel cycle, any deuterium-tritium fusion reactor must breed its tritium fuel onsite. Lead-lithium (PbLi), a eutectic metal, is a leading liquid breeder material for tritium generation. One challenge with PbLi blanket technology is the extraction of tritium from the molten eutectic. Three technologies are the focus of worldwide research: the vacuum permeator, the vacuum sieve tray, and the gas-liquid contactor (GLC). The present work offers a methodology for designing, sizing, optimizing, and costing a trickle-bed GLC. Here, we analyzed tritium extraction from PbLi using MELODIE experimental data by applying traditional packed bed mass transfer efficiency models along with supplementary models, like film theory. Our analysis revealed that traditional packed bed mass transfer efficiency models do not fit the MELODIE loop experimental data. Moreover, uncertainty in PbLi solubility resulted in a 325-fold increase in required gas flow rates when comparing identical packing heights. The film theory liquid mass transfer coefficient, Delt-Olujic wettability model, and Reiter tritium solubility values fit the MELODIE data best and were used both in the design and to conduct the economic analysis. Techno-economic analysis of the GLC was performed to evaluate three design sizes, all achieving a minimum extraction efficiency of 90 [%] for a total tritium extraction of 31 [kg/yr].

Fusion Fuel Cycle↗
DOE OSTI2024

Advancing Tritium Self-Sufficiency in Fusion Power Plants: Insights from the BABY Experiment

Abstract In the pursuit of fusion power, achieving tritium self-sufficiency stands as a pivotal challenge.
Tritium breeding within molten salts is a critical aspect of next-generation fusion reactors, yet experimental measurements of \gls{tbr} have remained elusive.
Here we present the results of the \gls{baby} experiment, which represents a pioneering effort in tritium research by utilizing high-energy (\SI{14}{\mega\electronvolt}) neutron irradiation of molten salts, a departure from conventional low-energy neutron approaches. 
Using a small-scale (\SI{100}{\milli\litre}) molten salt tritium breeding setup, we not only simulated, but also directly measured a \gls{tbr} \textcolor{red}{(\num{3.57e-4})}. 
This innovative approach provides crucial experimental validation, offering insights unattainable through simulation alone.
Moreover, our findings reveal a surprising outcome: tritium was predominantly collected as HT, contrary to the expected TF. 
This underscores the complexity of tritium behavior in molten salts, highlighting the need for further investigation.
This work lays the foundation for a more sophisticated experimental setup, including increasing the volume of the breeder, enhancing neutron detection, and refining tritium collection systems.
Such improvements are crucial for advancing our understanding of fusion reactor feasibility and paving the way for future experiments.

Delaporte-Mathurin, Rémi (ORCID:0000000310648882)↗
DOE OSTI2025

Tritium Trapping Thermodynamics by Point Defects at Interfacial Fe-Al Phases of the Aluminide Coating

Density functional theory (DFT) simulations have been carried out to evaluate the potential for tritium trapping by metal vacancies in four phases (i.e., (Fe, Cr, Ni), AlFe 3 , NiAl, and AlFe) identified near the interface between the Al coating and 316 stainless steel (316 SS) cladding. In addition, an ab initio thermodynamics approach has been employed to predict the temperature and T 2 -partial pressure dependence on the thermodynamics of tritiated defects. Key results in this work suggest that metal vacancies in the four phases have the potential to favorably trap tritium species. This thermodynamic trend can be correlated to the high energy cost of having an interstitial tritium in the lattice, which has been calculated to be at least 0.48 eV. By combining the results of this study with previous theoretical works investigating tritium behavior in other Fe-Al phases identified in the aluminide coating, suggest that metal vacancies are generally able to trap tritium in various Fe-Al aluminide phases. Especially, it was found that Al vacancies are the most efficient to trap tritium, followed by Fe vacancies, then Ni and Cr vacancies. Strong interactions between tritium and the metal vacancies are occurring by the formation of Fe—T bonds. The formation of Cr—T, Ni—T, or Al—T bonds are found less energetically favorable than Fe—T bonds.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Thermodynamics of Tritium Trapping by Point Defects at Interfacial Cr-based Phases of the Al Coating

Density functional theory (DFT) simulations have been carried out to evaluate the potential for tritium trapping by metal vacancies in four phases Cr-containing phases (i.e., Cr3Si, Al0.3Cr0.7, Al8Cr5-HT, and Al8Cr5) identified near the interface between the Al coating and 316 stainless steel (316 SS) cladding. In addition, an ab initio thermodynamics approach has been employed to predict the temperature and T2-partial pressure dependence on the thermodynamics of singly tritiated defects. Key results in this work suggest that metal vacancies in the four phases have the potential to favorably trap tritium species, especially Si vacancies in Cr3Si phase. This overall thermodynamic trend can be correlated to the energy cost of having an interstitial tritium in the lattice, which has been calculated to range from 0.17 eV in Al8Cr5-HT to 0.89 eV in Cr3Si. A comparison of the results from this study with previous theoretical works investigating tritium behavior in other Fe-Al phases identified in the aluminide coating, suggests that metal vacancies are generally able to trap tritium in various Fe-Al aluminide phases. Especially, it was found that Si and Al vacancies would be the most efficient to trap for tritium, followed by Cr vacancies, then Fe and Ni vacancies. In the four Cr-based material phases investigated in this work, it is interesting to note that, in the absences of Fe or Ni species, strong interactions between tritium and the metal vacancies are always occurring by the formation of preferential Cr—T bonds (i.e., no Si—T or Al—T bonds were formed).

Sassi, Michel J.↗
DOE OSTI2025

The molten salt tritium transport experiment: A pumped fluoride salt loop for hydrogen isotope experimentation

Molten salt reactors (MSRs) and fusion reactors propose to use molten salts as coolants and breeder blanket materials, respectively. Tritium, however, poses safety concerns in both reactor types due to its ability to permeate through reactor materials and potential for environmental release. This manuscript addresses the tritium transport phenomena in molten salts and presents the design and analysis of the Molten Salt Tritium Transport Experiment (MSTTE). MSTTE is a forced-convection fluoride salt loop intended to measure hydrogen isotope permeation through structural materials in a flowing salt system. In the first phase, MSTTE will use FLiNaK salt and deuterium as surrogates for FLiBe and tritium, with future plans to utilize tritium and FLiBe. MSTTE couples a Copenhagen Atomics pumped salt loop with an external test section that introduces hydrogen isotopes into the loop and measures transport phenomena. The Hydrogen Injection System (HIS) controls hydrogen isotope introduction into the molten salt loop. Here, the permeation test section measures the permeation rate through stainless steel tubing in contact with flowing salt. Computational fluid dynamics (CFD) analysis ensures fully developed salt flow in the permeation test section. MSTTE is modeled with MELCOR-TMAP to predict the permeation rate as a function of experimental variables such as source term, salt flow rate, and salt temperature. Results indicate that the source term is the only parameter with a significant effect on the permeation rate. Pressure drop analysis suggests that the loop should operate below 200 LPM to maintain a pressure drop below 200 kPa. Additionally, finite-element analysis assesses thermal stress during loop operation to ensure the experiment's safe design. MSTTE will provide semi-integral data on tritium transport phenomena in molten salts and serve as a testbed for advancing molten salt technology.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Species Transport Framework Development in SAM for System-Level Tritium Source Term Analysis

The SAM code is under development as a modern system-level modeling and simulation tool for advanced non–light water reactor safety analyses, with recent efforts to add capabilities to evaluate radiological source term risks in these novel reactor concepts. By leveraging the established system-level multiphysics thermal-hydraulic models in SAM, a framework for tightly coupled species transport modeling has been integrated into the code for engineering-scale source term evaluation. This species transport framework was first applied to the simulation of tritium, which is a well-known source term in conventional light water reactors. Tritium poses a unique risk in salt-cooled reactors, especially those with lithium-bearing salts such as the fluoride salt–cooled high-temperature reactor (FHR) concept, as tritium is generated in the salt coolant in significant quantities due to neutron interactions. A compounding factor is the increased mobility of tritium at high temperatures, which is able to permeate through metals while also potentially being retained in graphite pebbles and structures. Engineering-scale models for the tritium transport pathways in a FHR have been developed using the new species transport framework in SAM. The capabilities are assessed through analytical verification problems and validated with data from a graphite retention experiment. In conclusion, the system-level model is demonstrated by performing an initial estimate of baseline tritium generation and flows in a generic reference SAM FHR model, setting a foundation for future studies of source term transient analysis with the potential for further multiscale and multiphysics integration.

SAM↗