AVS 72 Session FUS-WeM: Fusion Materials, Plasma-Facing Components, and Tritium Engineering

Building on prior validation of a hydrogen-based surrogate unit for water detritiation, this workadvances catalyst development through integrated computational screening and targeted synthesis. The previous study established baseline performance using commercial Cu-Zn catalysts, revealing reaction rates limited by the internal mass diffusion. To identify candidates with potentially improved characteristics, Machine Learning (ML)-accelerated adsorption energy calculations (AdsorbML) were employed to screen a broader alloy design space. A promising composition was selected based on predicted surface energetics favorable for water splitting reaction adsorbates. The candidate catalyst was synthesized and characterized using XRD and XPS to confirm phase composition and surface chemistry. Temperature-programmed reduction (TPR) was performed to determine appropriate activation conditions prior to reactivity testing. The catalyst was then tested in the surrogate reactor under conditions comparable to prior experiments. Preliminary results confirm catalytic activity for hydrogen evolution with oxygen retention. This work demonstrates a computationally-guided pathway for catalyst discovery for thermochemical water detritiation, contributing toward optimized tritium recovery systems for closing the fusion fuel cycle.

This material is based upon work supported by the Department of Energy [National Nuclear Security Administration] University of Rochester “National
Inertial Confinement Fusion Program” under Award Number(s) DE-NA0004144.

MAX-phase Ti₂AlC thin films were synthesized using PVD magnetron sputtering of a bilayer system that utilizes an Al precursor layer with a subsequent TiAlC layer, followed by vacuum annealing at 650°C for 3h up to 20h. Ex situ XRD of TiAlC only films (no Al precursor layer) reveal that annealing drives Al out of the Ti_0.66Al_0.34C rocksalt phase present in the as-deposited films, generating additional C-vacancy sites to form TiC_0.5. No Ti₂AlC was synthesized from these TiAlC only films. With the utilization of an Al precursor layer deposited prior to the TiAlC layer, it was found that the as-deposited bilayer contains spatially distinct TiC_0.56 and Al phases, establishing the diffusion geometry required for Al intercalation-driven Ti₂AlC formation. This mechanism is initiated via vacuum annealing at 650°C which mobilizes Al both from the Ti_0.66Al_0.34C phase in the TiAlC layer but also Al from the precursor layer. The resulting upward Al diffusion and intercalation into ordered C-vacancy planes of TiC_0.55 is identified as the primary mechanism driving low-temperature Ti₂AlC formation in Al/TiAlC films. XRD results also show the formation of Ti₃Al along with the TiC_0.55 and Ti₂AlC phases, mirroring reported two-step reaction pathways that include the formation of Ti-Al intermetallics and TiC_x that react to form Ti₂AlC. Ex situ XPS depth profiles corroborate this mechanism, showing progressive Al depletion from the precursor layer into the TiAlC layer with annealing time. Associated scans of the Al 2p orbital taken during depth profiles shows binding energy shifts evidencing changes in the Al chemical environment consistent with C-vacancy incorporation and Ti₂AlC synthesis.

These efforts towards low-temperature synthesis of MAX-phase coatings allow for their potential use on low activation structural materials used in fusion reactor blanket components. Ti₂AlC is a particularly promising candidate for this application, as aluminum-containing MAX-phases form a dense, protective Al₂O₃ scale upon oxidation that has been demonstrated to reduce hydrogen isotope permeation through structural materials.

Due to their high permeability for hydrogen isotopes (HI), vanadium (V) membranes are being considered for separation of HI from the plasma exhaust of future fusion reactors. Surface effects, particularly the presence of impurities, have a large influence on the performance of such membranes. However, it is often difficult for conventional surface analysis tools to directly observe how HI interacts with other surface species. Ion scattering spectroscopy (ISS) and direct recoil spectroscopy (DRS) offer the advantage of being able to directly detect adsorbed HI by analyzing the energies of scattered and recoiled particles from ion scattering on the surface.

In this study, we use a 2 keV Ne⁺ ion beam to probe the surface composition of a polycrystalline V sample while dosing with hydrogen (H) or deuterium (D) with increasing pressure [Fig. S1(a)]. Gas species and surface temperatures (T_s) are varied for comparison, and a gas doser heated to 1100 °C is available for dissociating H₂/D₂ molecules into H/D atoms. DRS peaks of H/D are clearly identified in the ion energy spectra [Fig. S1(b)]. Pressure dependences of the peak intensities can be well fitted by a simple site-blocking model with Langmuir isotherm [Fig. S1(c)], which assumes that adsorbates only interact with each other by competing for empty adsorption sites. Such fittings reveal the change of the adsorption-desorption balance of HI with different surface conditions. The ratio between the adsorption rate coefficient and desorption rate coefficient, K = k_ad / k_de, decreases with increasing T_s, indicating that desorption is stronger at higher T_s [Fig. S1(d)]. K is also higher with atomic H/D dosing than with H₂/D₂ dosing, since atomic species have higher chemical energy to overcome the adsorption energy barrier, and thus result in higher adsorption rates.

In addition, preliminary studies of oxygen (O) adsorption by DRS while dosing the V surface with O₂ have been performed. The DRS O peak intensity growth with increasing O₂ pressure is similar to that measured by X-ray photoelectron spectroscopy (XPS) [Fig. S1 (e)], which validates the DRS results. The ongoing work includes sequentially dosing O₂ and HI to validate the site-blocking model. Finally, investigation of the effect of permeation barrier coatings, such as SiC, W or Al₂O₃, on V and V-alloys with both DRS measurements and density functional theory (DFT) modelling is also underway, and preliminary results from this work will be included in this presentation.

SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525

Metal foil pumps (MFPs) exploit plasma–driven hydrogen super–permeation to selectively separate and compress hydrogen isotopes (D/T) from helium and impurity species in the fusion exhaust stream, forming a key enabling technology for Direct Internal Recycling (DIR)–based fuel–cycle architectures. Despite their promise, the technical readiness level of MFPs remains insufficient for integration into next–generation fusion systems. Advancing the understanding of hydrogen super–permeation in metallic membranes is therefore essential for developing high–performance pumping components capable of operating under reactor–relevant plasma conditions.

This work presents an integrated experimental and theoretical investigation of deuterium super–permeation in high–purity α–Fe. Plasma–driven permeation experiments conducted in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory demonstrated that a 99.99% α–Fe foil can achieve a maximum pumping throughput of 3.0 L s⁻¹ cm⁻² and a pumping speed of 3.0 m³ s⁻¹ m⁻² at 523 K under a D₂ pressure of 0.013 Pa. The permeation behavior was found to be strongly governed by the surface sticking coefficient, α, and by the incident D⁺ ion energy. Comparison with the Pick–Sonnenberg model showed excellent agreement when using a previously measured α ≈ 1.6 × 10⁻⁹, with peak pumping performance occurring near an incident ion energy of 10 eV [1]. In contrast, incident energies above 50 eV caused a sharp decline in pumping performance, consistent with sputtering–induced removal of the native oxide monolayer—an effect also reported in prior studies of Nb and Fe.

A complementary theoretical study applied the Pick–Sonnenberg formulation to evaluate hydrogen permeation in candidate group–V membrane materials, focusing on vanadium with an experimentally measured effective permeability influenced by surface oxide and subsurface carbide layers [1]. Modeling showed that even with the reduced permeability, an uncoated 1 mm–thick V membrane can achieve approximately 10% hydrogen permeation at 973 K when operating with α ≈ 10⁻³ and ion fluxes near 10²⁰ m⁻² s⁻¹. These findings highlight the potential for high–temperature operation to enable attractive pumping performance without the Pd coatings typically required for group–V metals.

Collectively, these results deepen the mechanistic understanding of hydrogen super–permeation in metallic membranes and provide a physics–grounded basis for optimizing high–flux MFP systems within advanced fusion fuel–cycle concepts.

Reference:

[1]M.A. Pick and K. Sonnenberg, J. Nucl. Mater. 131 (1985) 208-220

While Tungsten (W) is a leading candidate for a plasma-facing material in fusion reactors, neutron-induced displacement damage creates vacancy-type defects that significantly increase tritium retention [1,2]. To mitigate this, we developed W alloys using undersized solute elements like chromium (Cr) and rhenium (Re). These elements promote the annihilation of vacancy-type defects during irradiation at temperatures ≥773 K and reduce the resulting trap density [3-8]. Furthermore, recent findings show these alloying elements enhance defect recovery during post-irradiation annealing at approximately 1473 K, offering a potential pathway for in-vessel trap removal during scheduled maintenance. This presentation explores the mechanisms driving these enhanced annihilation and recovery processes.

[1]Y. Hatano et al., Nucl. Fusion, 53 (2013) 073006.

[2]Y. Hatano et al., J. Nucl. Mater., 438 (2013) S114.

[3]Y. Hatano et al., Nucl. Mater. Energy, 9 (2016) 93.

[4]J. Wang et al., J. Nucl. Mater., 545 (2021) 152749.

[5]J. Wang et al., J. Nucl. Mater., 559 (2022) 153449.

[6]J. Wang et al., Mater. Design, 229 (2023) 111899.

[7]J. Wang et al., Mater. Design, 259 (2025) 114783.

[8]Y. Nobuta et al., J. Nucl. Mater., 566 (2022) 153774.

Glow discharge polymer (GDP) is the primary material used as an ablator for spherical, direct-drive laser experiments at the University of Rochester’s Laboratory for Laser Energetics. GDP capsules are millimeter-scale hollow spheres with gas permeation time constants of ~1 min for hydrogen isotopes—the gases used for fusion reactions. Gas permeation time constants, however, must be significantly longer (tens of hours) to meet the specified fuel pressure at the time of the laser shot. Most experiments require fuel pressures of 15 atm±10%. Sputter-deposited aluminum has been demonstrated to increase the gas permeation time constant, using layers of around 100 nm. However, sputtered aluminum films often exhibit a broad distribution of time constants, rendering many samples unusable. An alternative to sputter deposition is atomic layer deposition (ALD), where layers of Al₂O₃ are applied using a vapor-phase deposition process to produce a contiguous gas-permeation layer. ALD is an alternative method to increase hydrogen gas retention and is investigated in this study.

The presented work will report gas-permeation time constants on spherical and cylindrical plastic shells from a decay curve measured using a residual gas analyzer. In our time-constant measurement system, the targets are pumped out to high vacuum (10^–6 Torr) and filled to pressures up to 1000 Torr. Preliminary observations show time constants ranging from 1000s to 2000s, with a dependence on pressure. Changes in the permeation rate as a function of cycling will be investigated. Characterization (using x-ray photoelectron spectroscopy, a scanning electron microscope, and an atomic force microscope) will be performed to observe microscopic changes on the sample surfaces that impact gas retention.

*This material is based upon work supported by the Department of Energy (National Nuclear Security Administration] University of Rochester “National Inertial Confinement Fusion Program” under Award Number DE-NA0004144

Presentation Type: Oral presentation

Presenter: Mark Bonino, mbon@lle.rochester.edu [mailto:mbon@lle.rochester.edu], (585)275-1404 (office)

In order to gauge the impurities entering the tritium fuel cycle (FC) in an ICF target chamber, we focus on residual target material left after a shot in the forms of ablation coating and gaseous species. The ablation coating, or material left on surface of chamber, was measured by placing sapphire wafers 26 cm away from the target. The gaseous species are measured via Residual Gas Analyzer (RGA) directly mounted on the ICF target chamber. The RGA provides mass information for the major gas species residual in the chamber. After being exposed to an ICF target implosion, the wafers will be separated into two testing groups: one to measure tritium activity and the other to measure surface changes. Tritium surface activity will be measured through etching, while surface changes will be measured by X-ray photoelectron spectroscopy and X-ray reflectivity. The goal is to understand the depth that tritium reaches, surface changes on the wafer, contaminants left on surface, and contaminants left in gas.

This material is based upon work supported by the Department of Energy [National Nuclear Security Administration] University of Rochester “National Inertial Confinement Fusion Program” under Award Number(s) DE-NA0004144; Data were produced by Battelle Savannah River Alliance, LLC under Contract No.89303321CEM000080 with the Department of Energy.

Oak Ridge National Laboratory has partnered with Type One Energy and The University of Tennessee Knoxville to build a world-class High Heat Flux Experimental facility focused on testing and qualifying fusion energy plasma-facing components (PFCs). This facility will utilize two converging high-power electron beams (~1.6 MW total) to heat test articles at fusion-relevant PFC power densities. Heat will be removed from the test articles via a dedicated helium loop with flowrates up to 1 kg/sec at a pressure of 8 MPa. Large components measuring over 1 m x 1 m x 0.5 m can be tested in this facility, enabling sub-component PFCs to be evaluated under representative operating conditions for thermomechanical stress testing and qualification. There are currently no dedicated testing facilities in the Western Hemisphere with these capabilities. Therefore, the FORGED-PFC facility will fill this needed capability and operate as a community resource to accelerate domestic fusion research and deployment of fusion energy systems.

For effective high heat flux testing at this scale, experiments must take place in a vacuum environment at pressures on the order of 1x10^-4 mbar and lower. The two electron beams will be mounted to a relatively large vacuum vessel (~ 10 m³) that will also house the test articles during experiments. Robust thermal management of the vacuum vessel is required to dissipate up to 1.6 MW of heat from the inner walls, necessitating a jacketed design. In addition, numerous ports are required for diagnostics to enable reactor relevant measurements and support predictive modeling validation. To achieve target pressures and minimize downtime between experiments, a multi-stage roughing pump system is required in tandem with cryopumps. To optimize the layout of the electron beams and diagnostics relative to the test articles, a custom D-shaped vacuum vessel is envisioned for this application.

An overview of the FORGED-PFC project and its current status will be provided, along with design details of this high-performance vacuum vessel and associated pumping system.

Note: This work is supported by the U.S. Department of Energy contract DE-AC05-00OR22725. The views and opinions expressed herein do not necessarily reflect those of Oak Ridge National Laboratory.

The demonstration of laser-based inertial confinement fusion (ICF) ignition on the National Ignition Facility (NIF) in 2022 was a triumph of integrated physics and decadal science. This experiment, and the subsequent experiments building up to energy gains >4, are to date the only experimental demonstrations of fusion exceeding the Lawson criterion in the laboratory. Yet underlying this physics result was a complex blend of materials science and vacuum deposition needed to create and field an igniting target. This presentation will cover two aspects of the igniting physics package where vacuum deposition plays a key role: the hollow, spherical fuel capsule that implodes and the cylindrical hohlraum that provides x-ray drive for the implosion.

The capsule used in ignition experiments is CVD-deposited nanocrystalline diamond deposited. This talk will describe improvements to fabrication including machine-learning based diagnostics, Monte Carlo modeling, integrated experiments, and infrared reflectometry to improve coating thickness uniformity to > 99.8% on non-planar, non-stationary substrates.X-ray computed tomography was used to confirm that improvements to coating uniformity did not create additional internal defects.Finally, by controlling the composition along the growth axis we created capsules that stabilized the implosion and resulted in a record gain of >4. We will briefly mention materials science challenges of other capsule materials including boron carbide and glow discharge polymer.

The cylindrical hohlraum that contains the imploding capsule is a high-atomic number material (gold and uranium) that converts incoming laser light into a spherical bath of x-rays that drives the capsule implosion.Despite decades of study, the complex oxidation pathways of uranium are incompletely understood.We observed catastrophic oxide spalling on depleted uranium hohlraums that requires both an oxidant during processing and an environmental trigger. An experimental study confirmed that removing one or the other effectively increased the shelf-life of these components from ~weeks to ~months or years, eliminating a major cause of late-stage target failures for NIF.

LLNL-ABS-2019160. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344 and partially within the LDRD program under project 24-ERD-084.