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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

DOE Challenges and Opportunities Associated with Accountable Nuclear Material Needs for Development and Commercialization of Fusion Nuclear Energy

Fusion energy represents a transformative opportunity to deliver a safe, plentiful, and carbon-free source of reliable primary power. In recent years, fusion research and development have accelerated significantly, particularly within the US, driven by decades of foundational public investment. Notably, in 2024 the US Department of Energy (DOE) established a comprehensive Fusion Energy Strategy aimed at collaborating with industry partners to enable the deployment of fusion power plants and grid integration by the 2030s. This strategy is chiefly implemented through the DOE Office of Science (SC) Fusion Energy Sciences program. This project was initiated to identify potential approaches for the Office of Environment, Safety, and Health (NA-ESH-12) within DOE’s National Nuclear Security Administration to begin engagement with the fusion community on future accountable material needs. The goal of the project is to inform and influence the supply of and demand for accountable nuclear materials as fusion energy is developed and commercialized. NA-ESH-12 must proactively engage with the fusion community regarding the production and management of accountable nuclear materials. Given the complexity and scale of materials required for research, pilot projects, and eventual commercial reactors, early coordination is vital. The project’s objective is to provide insights that will shape the supply and demand landscape for critical nuclear materials, ensuring that DOE is prepared to effectively support fusion energy development and commercialization. In FY 2025, an initial limited review was conducted to identify the status of the fusion energy community’s progress toward full-scale energy production and the need for accountable nuclear material. This included communications with SC, NA-ESH-12, Savannah River National Laboratory, and Oak Ridge National Laboratory, and attending the Rutgers University–sponsored Supply Chain Workshop: Scaling the Fusion Industry and the International Atomic Energy Agency’s Ninth DEMO Programme Workshop. The review to date indicates that the amounts of tritium, lithium-6, and deuterium required by the fusion industry will be dependent on fuel type, breeding technology, blankets, and R&D improvements. One concern is that the commercial sector does not have a sufficient supply chain to meet the demand for development and commercialization for fusion energy production. The supply and demand estimates for these materials should be routinely reviewed as fusion technologies mature.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Materials Assessment for FLiBe Fusion Blankets

This evaluation was prepared at the request of DOE-FES and expands on an earlier evaluation prepared by Wilson, Wiffen and Keiser in 2015 (attached as Appendices A and B). Additional useful information for materials in fission molten salt reactor systems that use FLiBe can be found in a prior NRC report [Busby 2019]. The purpose of this report is to review the considerations that will determine the fusion blanket structural materials that will be suitable for use in a FLiBe breeder-coolant fusion power plant (FPP). Identifying a fusion-relevant structural material that is sufficiently compatible with FLiBe is a key issue to be addressed. The constraints and trade-offs made in structural materials selection include the usual power generation requirements such as mechanical properties, fabricability and durability, which assumes reasonable compatibility with the working fluid. For an FPP, additional constraints include neutron economy (breeding ratio), resistance to irradiation damage and low activation (i.e. waste disposal) characteristics.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Assessment of BISON capabilities for component-level prediction of tritium transport in fusion and fission applications

Here we assessed the hydrogen migration and redistribution model implemented in the BISON fuel performance code for component-level tritium transport applications in fusion and fission example scenarios. We developed BISON models of an ITER heat exchanger, light water reactor fuel cladding, and a fluoride salt cooled high temperature reactor heat exchanger. Reasonable agreement between reported values in reference studies and BISON predictions demonstrated the ability of the BISON models to predict tritium transport behavior through steel materials. Next, sensitivity and uncertainty analysis was used to understand key parameter sensitivities within the BISON model for evaluation of tritium migration and redistribution. . The sensitivity analysis showed that the diffusivity activation energy was the most important parameter and that the heat of transport was the least important parameter for tritium migration in steels. Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Fusion Fuel Cycle Technology Downselection Methodology

Tritium Start-up Inventory Requirements Current consensus has full inventory coming from CANDU reactors. • Assumes continued operation through 2035, yielding ~30 kg tritium. • Produced in the form of DTO, which requires separation for tritium extraction. • ITER will take between 10-20 kg. • DEMO will require ~15 kg. • Fuel cycle design requires minimum volume to fill plumbing in system beyond inventory necessary to reach break even. • With ~45 current private ventures, need massive uptick in tritium production to meet global need.

Malone, Collin↗
DOE OSTI2020

Stability and Control of Burning Tokamak Plasmas with Resistive Walls (Second Year Report/ Final Report)

This research will focus on quantitative prediction of the stability, control, and equilibrium state evolution in toroidal burning plasmas. The stability of long pulse burning plasmas is highly sensitive to the physics of resonant layers in the plasma, sources of momentum and ow, kinetic effects of energetic particles, and boundary conditions at the wall, including feedback control and error fields. In ITER in particular, the low toroidal ow equilibrium state, sustained primarily by energetic alpha particles from fusion reactions, will require the consideration of all of these key elements to predict quantitatively the stability and evolution. The principal investigators on this proposal are leading experts in the relevant theoretical and computational areas, and aim to perform computations guided by analytic modeling, to address this physics in realistic configurations. The overall goal is to understand the key physics mechanisms that describe resistive toroidal burning plasmas, surrounded by a resistive wall, under active feedback control. With the physics of the energetic ions, resonant layers, resistive wall, and toroidal momentum transport included, this study will extend from recent publications in theory and simulation of individual effects and move toward predictive modeling for burning plasmas.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Flexible Stellarator Physics Facility

We propose to build a Flexible Stellarator Physics Facility to explore promising regions of the vast parameter space of disruption-free stellarator solutions for Fusion Pilot Plants (FPPs). The FESAC Long Range Plan recognized the quasi-symmetric stellarator as “the leading US approach to developing disruption-free, low-recirculating-power fusion configurations”. To deliver the ambitious Decadal Vision for Commercial Fusion Energy, we must establish a persuasive stellarator program in parallel to the tokamak one: the stellarator will prove to be a better path to a reactor if theoretical predictions are confirmed and novel optimization techniques and strategies work as desired. Since the release of the Long Range Plan in 2020, stellarators have arguably made the most significant advances of all fusion concepts. Groundbreaking results from W7-X demonstrated low neoclassical transport and the successful operation of the island divertor. Advances in theory and modeling now allow us to minimize turbulent transport, to achieve equilibria with precise quasisymmetry, to reduce neoclassical transport and fast ion loss to levels far below what has been previously achieved, and to minimize the effect of coil manufacturing errors. If realized, these advances will lead to cost-effective stellarator designs with confinement comparable to tokamaks but without the fundamental challenges of disruptions and current drive.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Understanding the Quadrupole Mass Filter and Testing a High-Resolution QMS RGA for ITER

A common type of residual gas analyzer is the quadrupole mass spectrometer. One of the main components within this instrument is a mass filter known as the quadrupole. It is responsible for the selective throughput of the ionized gas particles - by ascending mass number - prior to ion impacts on the analyzer (or detector) surface from which the ion current signal is generated for processing. However, the quadrupole is not fully described in relation to the electric field characteristics and the function as an ion mass separator. This paper describes the basic origins of the electrical design, the intricate assembly criteria, and performance of the quadrupole within the spectrometer. A specialized quadrupole mass spectrometer is part of a configuration for a diagnostic gas analyzer system planned for ITER, a fusion research machine. It has a verified capability, essential as a diagnostic criterion for this reactor project, to successfully deconvolute the mass signals of Helium-4 and deuterium (reactor fuel exhaust gases, separated by only 0.026 atomic mass units), down to a relative three-percent concentration of the former gas. The associated preliminary testing, performed at the Oak Ridge National Laboratory, is also addressed. Finally, one of the key parameters used to express gas concentration, the relative sensitivity factor, will be explained, including an evaluation of dependency on other variables.

Marcus, Chris [ORNL] (ORCID:0000000190139636)↗
DOE OSTI2024

Characterization of the ELM-free negative triangularity edge on DIII-D

Tokamak plasmas with strong negative triangularity (NT) shaping typically exhibit fundamentally different edge behavior than conventional L-mode or H-mode plasmas. On DIII-D, every plasma with sufficiently negative triangularity (δ < δ crit ≃ -0.12) is found to be inherently free of edge localized modes (ELMs), even at injected powers well above the predicted L-H power threshold. It is also possible to access an ELM free state at weaker average triangularities provided that at least one of the two x-points is still sufficiently negative. Access to the ELM-free NT scenario is found to coincide with the closure of the second stability region for infinite-n ballooning modes, suggesting that ballooning stability may play a role in limiting the accessible pressure gradient in NT plasmas. Despite this, NT plasmas are able to support small pedestals and are typically characterized by an enhancement of edge pressure gradients beyond those found in traditional L mode plasmas. Further, the pressure gradient inside of this small pedestal is unusually steep, allowing access to high core performance that is competitive with other ELM-free regimes previously achieved on DIII-D. Since ELM-free operation in NT is linked directly to the magnetic geometry, NT fusion pilot plants are predicted to maintain advantageous edge conditions even in burning plasma regimes, potentially eliminating reactor core-integration issues caused by ELMs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Preliminary neutronic and thermal analysis of Inertial Fusion Energy systems with a focus on thick liquid wall concepts

This paper provides an assessment of the neutronic behavior of Inertial Fusion Energy systems using the Serpent 2 Monte Carlo code. The study explores different types of reactor concepts, including dry-wall and thick liquid wall systems. In particular, it characterizes the tritium breeding and energy multiplication performance, as well as the effect of long-term irradiation on structural materials. For this purpose, the displacement per atom (DPA) and the gas production rate in structural materials are calculated. A simplified geometry with varying geometrical parameters is used as a representation for the different kinds of devices. Additionally, the case of the HYLIFE-II concept is used to assess modeling inaccuracies arising from certain approximations. A comparison between the simplified geometry and a realistic one is performed, thanks to Serpent’s capability to run transport calculations on unstructured meshes. The study then addresses the impact of these discrepancies on the temperature distribution in the first wall using the GeN-Foam code for conjugate heat transfer analysis. The results highlight the necessity to consider the heterogeneous nature of the liquid wall, as simplified models tend to underestimate radiation damage metrics and heat loads.

Conjugate heat transfer↗
DOE OSTI2020

Interactions between atmospheric pressure plasmas and metallic catalyst particles in packed bed reactors

Atmospheric-pressure plasmas sustained in packed bed reactors (PBRs) are being investigated for chemical conversion of gases and pollution control. Metallic catalysts added to the surfaces of the dielectric beads of PBRs can increase the energy efficiency and selectivity of chemical processes by reducing operating temperature and providing additional reaction pathways. In this paper, results from a computational investigation of plasma surface interactions between micron-scale metallic catalysts and humid-air plasmas in PBRs are discussed. Here we found that high plasma density regions form in the proximity of the metallic catalysts. These higher-density plasma regions were confirmed experimentally using ICCD imaging. The intense plasmas result from geometrical electric field enhancement and redistribution of charges within the conductive particles, leading to further enhancement. The high electric field at the triple points of the catalysts can produce electric field emission of electrons, which provides a pre-ionization source or additional source of electrons. These regions of high electric field and sources of electrons guide discharges towards the catalysts and increases fluxes of excited species, ions, electrons and photons to their surfaces. These fluxes are focused primarily at the triple points between the metal, dielectric and gas. As a result, the catalyst is locally heated, which could lead to further increased rates of thermocatalytic reactions on the surface. Surface roughness of the metal inclusions can lead to additional electric field enhancement, which changes the character of the discharges in the vicinity of the catalysts while reducing breakdown voltage.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Deuterium permeation and retention in 316L Stainless Steel Manufactured by Laser Powder Bed Fusion

Deployment of additively manufactured materials into nuclear energy systems requires investigation of the full range of the unique environmental effects on these materials. Hydrogen isotopes are common gaseous species in nuclear reactors and exhibit ample high mobility in most materials. While hydrogen isotopes mobility in wrought stainless steel is well understood, this property is not thoroughly studied for its additively manufactured variants. In this study, we investigated the deuterium permeation and retention in 316L stainless steel manufactured by laser powder bed fusion. The results showed that the deuterium permeability in the as-built additively manufactured 316L stainless steel (AM SS316L) is greater than that of the reference wrought 316L stainless steel by a factor of 2.8. However, the stress-relieved and solution-annealed AM SS316L samples exhibit lower deuterium permeability in comparison with the as-built condition. Following the solution annealing, the deuterium permeability of AM SS316L is comparable with that of the reference wrought SS316L. Deuterium retention in the as-built AM SS316L is 45% higher than that of the wrought 316L stainless steel and slightly higher than that of the two thermally annealed AM SS316L materials. Finally, transmission electron microscopy and positron annihilation lifetime spectroscopy were used to obtain microstructural information used to determine deuterium permeation and retention behavior in the studied materials.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Electrochemical hydrogen isotope fractionation - fundamental insights leading to process scale up

Introduction Deuterium and tritium, the two isotopes of hydrogen besides protium, are utilized in many industries. Deuterium is routinely used as a nuclear reactor moderator and molecular labeling agent, while tritium is used as an autoradiography label in medical imaging, and in an emerging market for self-powered lighting for watches, exit signs, firearm night-sights and future fusion fuel. Tritium is produced naturally at very low steady-state levels due to cosmic rays, natural nuclear decay reactions, as well as commercial nuclear fission reactors, and terrestrial nuclear testing conducted primarily in the mid-twentieth century. Tritium often accumulates in the heavy water used in nuclear fission reactors and sometimes appears in groundwater near nuclear reactors due to slight leakage from water sources in the reactor. Because of the health and environmental impacts from tritium leaking into the water sources, as well as the value of this precious material, it is desirable to remove and recover as much tritium as reasonably possible from contaminated water.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Tokamak level performance in the optimized stellarator Wendelstein 7-X with stable peaked density profiles

Progress towards achieving net gain from nuclear fusion is typically characterized by the triple product 𝑛⁢𝑇⁢𝜏. Although stellarators have several benefits for a reactor, such as the inherent steady-state capability and the lack of disruptions, the maximum achieved triple product has historically lagged behind that of tokamaks. In the large optimized stellarator Wendelstein 7-X, a triple product of (1.10 ± 0.15) × 10 20 m −3 ⁢ keVs has recently been achieved and held stable for 1.9 s. This marks an important step in stellarator core plasma performance and puts it in line with 𝐻-mode performance in similar sized tokamaks. The progress in Wendelstein 7-X performance is due to the suppression of ion-temperature-gradient turbulence by strong density gradients. In the record case, these are maintained by a fine balance of the core particle transport and central fueling using neutral beam injection. The duration of the scenario was limited in time only by technical restrictions of the neutral beam injectors and control of the divertor heat loads. This paper presents the result in the context of the leading magnetic confinement devices along with details of the scenario.

Bannmann, Sebastian [Max Planck Institute for Plas↗
DOE OSTI2024

Deuterium Permeation Through a Self-Supported Palladium-Silver Membrane in Helium Gas Mixtures

Palladium membranes and membrane reactors can separate and purify tritium from impurities in the plasma exhaust processing system for the fusion energy fuel cycle. Membranes can also act as a continuous separation method to remove tritium from helium streams in the breeder blanket tritium extraction system, such as from the purge gas of solid breeders. To investigate the potential of these membranes for these applications, we performed a deuterium permeation campaign with a self-supported palladium-silver (Pd-25Ag wt%) tube of 0.15 m length, 76 µ m wall thickness, and 3.0 mm inner diameter. A gas mixture of 3.95% D 2 and a balance of He was delivered to the inside of the tube and D permeated radially outwards through the membrane into a vacuum chamber. Permeation experiments were conducted at 300 ° C, 350 ° C, 400° C, and 450°C, from 100 to 1000 sccm total flow rate, and with total pressures of 90, 150, 190, and 250 kPa. Further, deuterium permeation was consistently lower than predicted from diffusion-limited permeation models, thus we developed a transport model that included gas-phase mass transfer and surface reactions to model experimental results. The dissociation constant was optimized to fit the developed model to experimental data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Divertor shaping with neutral baffling as a solution to the tokamak power exhaust challenge

Exhausting power from the hot fusion core to the plasma-facing components is one fusion energy’s biggest challenges. The MAST Upgrade tokamak uniquely integrates strong containment of neutrals within the exhaust area (divertor) with extreme divertor shaping capability. By systematically altering the divertor shape, this study shows the strongest evidence to date to our knowledge that long-legged divertors with a high magnetic field gradient (total flux expansion) deliver key power exhaust benefits without adversely impacting the hot fusion core. These benefits are already achieved with relatively modest geometry adjustments that are more feasible to integrate in reactor designs. Benefits include reduced target heat loads and improved access to, and stability of, a neutral gas buffer that ‘shields’ the target and enhances power exhaust (detachment). Analysis and model comparisons shows these benefits are obtained by combining multiple shaping aspects: long-legged divertors have expanded plasma-neutral interaction volume that drive reductions in particle and power loads, while total flux expansion enhances detachment access and stability. Containing the neutrals in the exhaust area with physical structures further augments these shaping benefits. These results demonstrate strategic variation in the divertor geometry and magnetic topology is a potential solution to one of fusion’s power exhaust challenge.

magnetically confined plasmas↗
DOE OSTI2020

Collaborative Research: Understanding Nanoparticle-Plasma Interactions in Dusty Non-Thermal Plasmas by Nanoparticle Probes and Online Aerosol Characterization (Final Technical Report)

This report describes results from the University of Minnesota’s contribution to studies performed under the award DE-SC0018202 (principal Investigator: Christopher J. Hogan Jr.. This award is a collaborative research grant with support equally provided to Case Western Reserve University and to the University of Illinois (with principal investigator R. Mohan Sankaran). To date ten peer-reviewed published papers have resulted from the studies performed with University of Minnesota personnel contributing; this report describes the results of these studies as well as several unpublished studies, grouped into seven separate study types. The overarching goal of the carried out studies was to understand nanoparticle behavior and dynamics in non-thermal plasma synthesis reactors, specifically adapting aerosol measurement technology to provide online measurements of particle size, charge, and structure in plasma reactor systems.

42 ENGINEERING↗
DOE OSTI2025

Uncertainty quantification in MELCOR Safety analysis of ARIES reactor designs

MELCOR-TMAP is a combined thermal-hydraulics and tritium tracking code developed to simulate severe accident scenarios in fission and fusion power plants. Here, we demonstrate the results of MELCOR-TMAP analyses on historical ARIES program reference designs. By coupling MELCOR-TMAP with the open source RAVEN probabilistic risk analysis framework’s Bayesian UQ capabilities, we also demonstrate key uncertainties in material properties with the highest impact on tritium inventory and plant risk.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗