Fusion Reactor Materials Research Articles

First-principles studies of behavior of hydrogen and mechanical properties of TiCrTaV high-entropy alloys

This study investigates the dissolution and diffusion of H atom in TiCrTaV of two body-centered cubic (BCC) high-entropy alloys and their effects on the mechanical properties and electronic structure of the alloys by first-principles calculations. Results show that the solution energy of H atom and local lattice distortion vary depending on the local environment. Meanwhile, the site occupancy of H atoms in tetrahedral interstitials are slightly advantageous in the two structures. Diffusion of H also presents a sluggish effect. The dramatic increases and decreases of diffusion barrier disrupted the periodicity and symmetry of H diffusion in the two BCC structures, thereby slowing down the H accumulation. After hydrogenation, TiCrTaV high-entropy alloys (HEAs) still maintain resistance to deformation, indicating that they are the potential candidate structural materials for fusion reactors. Our investigation indicates that this low-activation HEA may exhibit promising irradiation resistance due to its special H behavior.

Experimental results of deuterium/hydrogen co-current permeation through Nickel membrane, in non-steady-state surface limited regime

A better understanding of mutual influence among hydrogen isotope species in multi-isotope permeation is important for accurate assessments and predictions on tritium permeation through structural materials of fusion reactors. Moreover, this is important as the injection of H2 gas into the helium coolant of HCPB breeding blanket is proposed for mitigation of the permeation flux of tritium from the purge gas into the coolant. Although various modeling and experimental works have been carried out so far on this topic of multi-isotope permeation, unresolved issues or even controversies still persist. This work presents the results of some experimental tests of co-current permeation of deuterium with hydrogen through a very thin (0.075 mm) nickel membrane, performed in non-steady-state and surface limited regime (SLR), with the purposes to observe the effect of hydrogen on the permeation effective flux of deuterium, as well as to verify experimentally the predictions of a previous theoretical work. Thus, two deuterium/hydrogen gas mixtures of different ratios, as well as pure deuterium gas, were alternately introduced into the high-pressure side of the membrane by keeping the effective partial pressure of deuterium as close as possible to the same value (0.1 Pa), so that the comparison between multi- and mono-isotope permeation be relevant. The results indicate that the presence of (co-permeation) hydrogen does not lead to the decrease of deuterium permeation, but, on the contrary, to its increase. This effect is clearer revealed well before steady state is reached, where the effective permeation flux of deuterium has a constantly higher rise-rate when (co-permeation) hydrogen is present, and this rise-rate is even higher when the effective hydrogen pressure is higher. Also, a quite good consistency between experimental results and the theoretical predictions has been obtained, as well as a good repeatability of the results.

Inhomogeneous deformation substructure and its effect on the recrystallization behavior of yttria dispersion-strengthened tungsten plate

Tungsten materials have been chosen for use in critical parts (i.e., as the armor of plasma-facing material) in fusion reactors. Deformed tungsten will undergo recrystallization at a high-temperature service environment, and recrystallized tungsten exhibits brittleness, which reduces mechanical properties, thermal shock resistance and service lifetime. The deformation heterogeneity may result in an inhomogeneous microstructure, affecting its recrystallization behavior. The main aim of the present work is to qualitatively and quantitatively characterize the heterogeneity in the through-thickness deformed microstructure of an yttria dispersion-strengthened tungsten plate warm-rolled to 50 % thickness reduction ratio and study the effect of heterogeneous microstructure on recrystallization behavior. For this, topological maps of the surface, quarter and center region are measured on the deformed state and annealed state utilizing electron backscatter diffraction (EBSD). Distributions of grain size, geometrically necessary dislocations (GND) energy, potential nuclei density and texture have been analyzed in detail. The results revealed that the deformed microstructure of tungsten plate along full-thickness has no apparent heterogeneity gradient of grain size and texture, but the center layer has a larger subgrain size and more potential recrystallized nuclei compared to the surface layer. The larger subgrain is more conducive to static recrystallization nucleation, while the existing potential recrystallized nuclei can grow rapidly without incubation time during subsequent high-temperature annealing, thereby recrystallization preferentially undergoes in the center layer of the tungsten plate.

Mechanical properties evaluation of ReBCO CICC jacket based on super-austenitic stainless steel for CFETR high-field magnet

To excite stronger magnetic fields, high temperature superconducting (HTS) insert coils are planned for the next generation central solenoid magnet for China fusion engineering test reactor (CFETR). The development of high strength and high toughness jacket has become one of the challenges in the application of high-field cable-in-conduit conductors （CICC）. 0.2% proof stress of jacket should be over 1500 MPa at 4.2 K for HTS conductors, thus 316LN and JK2LB jackets developed by ITER do not meet the requirements. Nitronic 50 (N50) super-austenitic stainless steel material has great optimization potential for the development of jacket. Based on the modified N50 material, China prepared the CICC jacket and has done some R&D work on the N50 jacket. The entire process of ReBCO CICC preparation, including extrusion, bending and straightening, was simulated using the N50 stainless steel jacket. The N50 circle-in-square jackets prepared in China showed a yield strength greater than 1600 MPa and a fracture toughness KIC better than 110 MPa·m1/2 after cold work at 4.2 K. This paper will focus on the preparation, cold work processing and performance test of the N50 jacket. This study will present experimental data and discuss the feasibility of modified N50 as a high-magnetic field jacket material for next-generation fusion reactors.

Quantitative estimation of tritium amount in surface layer and bulk of tungsten sample using BIXS

Estimation of both the surface tritium and tritium in bulk in key materials of fusion reactors is of great importance for tritium safety and the management of tritium-contaminated material. For the further quantitative analysis of tritium in solids, the elaborate BIXS spectra were calculated based on Monte Carlo simulation. Four types of tritium depth profile were considered to evaluate the quantitative estimation method. It is found that the attenuation of X-rays in tungsten depends on both the energy of X-rays and the depth of X-rays. The evaluation of tritium amount in surface layer indicated that the intensity of Ar(Kα) peak could be used to evaluate the surface tritium within 400 nm from the surface in most cases and the deviations were less than 9% in the calculation. The intensity of both W(Lα) X-rays and the high energy X-rays can be employed to roughly estimate the total tritium amount. For linearly decreasing and exponentially decreasing distribution, the maximum calculation deviations were 24.9% and 28.8%, respectively. While for Gaussian distribution, the maximum deviations were 146% and 53%, respectively. And it can also be used for tritium estimation in other materials.

Interface alloying design to improve the stability and cohesion of W/HfC interface by first-principles study

Tungsten (W) and its composites have been regarded as the most promising candidate plasma-facing materials (PFMs) for fusion reactors due to its exceptional properties. In this work, we systematically investigated the effects of alloying elements (i.e., main groups, 3d-, 4d- and 5d-transition metals) on the stabilities and cohesion properties of the W/HfC interface by first-principles calculations. It is found that the alloying atoms of all main groups and transition metals of group VIB to IIB tend to segregate at the interface and increase the interface stability. In addition, when the alloy atom is doped, the change of interfacial bonding strength is highly correlated with the variation of bond length between interfacial atoms, and a linear model was proposed to account for most of the variations in the relative work of separation, with a R2 value of 0.94. Among these alloying elements, Re was found to enhance the cohesion strength of W/HfC interface by forming a strong Re-C bond based on the electronic structure analyses, and it is universal for Re to enhance the interfacial bonding strength in other W-transition metal carbide composites. This work provides a theoretical guidance for screening alloying elements that can improve mechanical properties of carbides dispersed W alloy in experiments.

Effect of irradiation temperature on radiation hardening of CLF-1 steel

A kind of reduced-activation ferritic-martensitic steel, CLF-1, was developed by the Southwestern Institute of Physics, China, as a candidate structural material for fusion and fourth-generation fission reactors. This work reports the results from the nanoindentation and transmission electron microscopy study performed on this steel after irradiation with 2.75 MeV Fe11+ ions up to 1 and 10 dpa at temperatures of 300 - 500 °C. An increase in hardness was observed in the damaged layer of samples irradiated at 400 °C, while the samples irradiated at 500 °C exhibited a softening of the damaged layer. The hardening can be primarily, although not exclusively, attributed to the increase of <100>-type dislocation loops observed at the increasing irradiation temperature. Nano-sized cavities around the dislocation loops and along dislocation lines were also observed, explaining the lattice swelling after irradiation at 500°C. The research results provide a valuable reference for the development of irradiation-hardening-resistant Fe-Cr alloys for advanced nuclear energy systems.

Helium irradiation induced microstructural damages and mechanical response of Al2O3-ZrO2-SiC composites

Ceramic composites are promising candidates as structural materials for future fission and fusion reactors. In present work, Al2O3-ZrO2-SiC ternary ceramic composites were irradiated with 2.0 MeV He-ions at 300 and 800 ℃. Grazing incidence X-ray diffraction results confirmed that there was irradiation induced shift and broadening of diffraction peaks, but no amorphization of Al2O3-ZrO2-SiC composite were observed up to fluence of 1.72 × 1018 ions/cm2. Transmission electron microscopy observations showed that throughout the entire irradiation region, nano-sized helium bubbles are mostly distributed in Al2O3 grains and partly in ZrO2 grains, while no detectable bubbles are observed in SiC grains. No obvious agglomeration of bubbles was found at grain/phase boundaries. By using nanoindentation technique, slight hardening or softening was confirmed for the samples irradiated at 300 and 800 ℃ respectively. The absence of amorphization and surface exfoliation indicating the Al2O3-ZrO2-SiC composite exhibits remarkable resistance to He-ions irradiation.

Decreasing the W-Cr solid solution decomposition rate: Theory, modelling and experimental verification

Tungsten-based alloys are essential materials in the nuclear fusion reactors’ development. Particularly, tungsten-chromium single solid solution is interesting for the combination of a good oxidation behaviour and lack of lower-melting-point phases. The W-Cr system has a miscibility gap though and the solid solution within the particular condition is inherently metastable/unstable and tends to decompose forming a Cr-rich phase. In this study, we have combined computer simulations and experimental work to find additional alloying elements to promote the stability of W-Cr solid solution. The results showed that alloying with Ta may result in lowering the stabilising temperature of the alloy, positively affecting the decomposition rate. Experiments performed on a W-Cr-Ta alloy showed prolonged solid solution lifetime when compared to W-Cr alloy. Formation of the third decomposed phase, enriched in Ta, was observed and changes in the chromium diffusion in the alloy were identified to be the governing factor for slowing down the decomposition.

Ab-initio investigation of mechanical and fracture-related properties of W-Re σ and χ precipitates

Tungsten (W) is a leading candidate for plasma-facing materials in fusion reactors. Recently, experiments have shown that during neutron irradiation, W and its transmutation products, mainly rhenium (Re), will form precipitates of the σ and χ types. This study identifies close-packed planes of the σ- and χ-phases in the W-Re system and uses ab-initio methods to identify mechanical properties such as elastic constants, generalized stacking fault energies (GSFE), and fracture toughness associated with brittle and ductile mechanisms. By utilizing a sublattice model we demonstrate how these properties depend on the Re content. For Re concentrations in the range where the σ- and χ-phases are stable, we find a weak dependency on the Re content, meaning that the elemental composition has little influence on the mechanical and fracture-related properties.

Materials for fusion reactors: Containing a star on Earth

Materials for fusion reactors: Containing a star on Earth Sandy (Alexander) Knowles, the Associate Professor in Nuclear Materials at the University of Birmingham, explores experimental metallurgy research and its structural materials to withstand the extreme temperatures inside nuclear fusion reactors. Fusion Energy provides near limitless energy without generating carbon emissions, or long-live radioactive waste. This entails it can be used for its energy density and small land usage, making it capable of providing stable base-load electricity, making it an ideal solution to our present energy challenges. Companies globally are currently seeking new approaches to fusion energy with development and fast deployment timelines, ultimately seeking to bring fusion to market sooner, potentially before 2040. Representing a ‘moonshot' innovation – with equivalence to the moon landing in terms of the pace and scale of developments needed, this collaborative societal grand challenge of fusion energy is a new opportunity for science, technology, engineering and mathematics (STEM), at all levels.

Microstructure and mechanical properties comparison of ferritic/martensitic steel fabricated by wrought and selective laser melting

Reduction activated ferrite/martensitic (RAFM) steel has been identified as the most promising structural material for future fusion reactors due to its properties such as low activation and high thermal conductivity. However, the complex internal water cooling tubing increases the difficulty of conventional processing. Herein, RAFM steel with high strength and excellent ductility was successfully fabricated by selective laser melting (SLM) technology. Compared with wrought materials, the SLM-fabricated specimens exhibit significant superiority in mechanical properties. Furthermore, the strength is prominently enhanced in the horizontal direction, with almost no loss of ductility. However, the strength gain in the vertical direction is at the expense of ductility. This work provides a potential method for the preparation of structural materials for nuclear fusion.

Comprehensive investigation of lattice dynamics of neutron multiplier beryllides: vibrational spectra and thermal performances

In order to realize tritium self-sufficiency during the operation of fusion reactors, neutron multiplier materials are needed to multiply the number of neutrons. Beryllium (Be) and beryllides (Be12V, Be12Ti) are considered as candidates for neutron multiplier materials in the tritium breeding blanket. Thermal performances are essential data to evaluate the service performance of materials in future fusion reactors. Due to the toxicity of beryllium, the preparation and analysis of neutron multiplier materials are challenging. As a result, the performance data of neutron multiplier materials are limited. It is essential to adopt the theoretical method to get the basic parameters for the design of a tritium breeding blanket and optimization of neutron multiplier. In the present work, the electron structure, vibrational spectra, and thermal performances have been theoretically investigated to understand thermal expansion and thermal conductivity properties. Experimental data are also included to compare with the calculated results and detailed analysis is presented. Due to the reactivity with water, BeO will form on the surface of Be. Therefore, BeO was studied as well. The bulk modulus of Be is the smallest of the four materials, while BeO is the largest. Higher bulk modulus has a better ability to restrain the deformation due to pressure. The c-axis of the Be12V has the largest thermal expansion coefficient, while the c-axis of the Be12Ti is the smallest. It has been found that the larger the lattice constant of crystal axes, the smaller the linear thermal expansion coefficient in same material. Electrons play a major role in the heat transfer of Be12V and Be12Ti. At high temperatures, phonon and electron coupling must be considered. Isotopeenrichment and enlarging the grain size cannot significantly improve the phonon thermal conductivity when the temperature is higher than room temperature. The calculated values can be adopted to predict the materials’ performances and support future fusion reactors’ neutron multiplier material preparation.

Analysis of dislocation characteristics and inside-outside contrasts in irradiated and annealed tungsten as a fusion reactor material

Tungsten is an important candidate of plasma-facing material for fusion reactors. Its irradiation response, especially the post-irradiation annealing (PIA) behavior needs further investigating. In addition, the practice of the “inside-outside” contrast method of determining the characteristics of irradiation induced dislocation loops has not been utilized frequently, and the present research serves as an example to present some practical considerations. In the present work, a tungsten thin-foil specimen is irradiated at 400 ℃ with 58 keV D&lt;sup&gt;+&lt;/sup&gt; to a final fluence of 1× 10&lt;sup&gt;17&lt;/sup&gt; cm&lt;sup&gt;–2&lt;/sup&gt;, corresponding to a dose of about 0.1 dpa. The specimen is prepared through the electro-polishing method by using a NaOH based electrolyte. The ion irradiation is carried out directly on the electro-polished specimen. The irradiated specimen is followed by isothermal annealing at 900 ℃ for 1 h. The as-irradiated and post-irradiation annealing modified defects are investigated with a transmission electron microscope (TEM) operated at 200 kV. The irradiation defects are characterized by using TEM bright-field imaging for the same imaging field with different &lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; vectors around the three major zone axes: the [001], [111], and [011] zone axis of the body-center cubic lattice of W. For each &lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; vector, the ±&lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; are characterized, and the corresponding contrast-extinctions and “inside-outside” contrasts of selected dislocation loops are identified. The indices of the &lt;i&gt; &lt;b&gt;g&lt;/b&gt; &lt;/i&gt; vectors around different zone axes are assigned consistently with the Kikuchi map. As a result, the D&lt;sup&gt;+&lt;/sup&gt; irradiation increases a fine distribution of dislocation loops with an average size of (11.10±5.41) nm and a bulk density of 2.40×10&lt;sup&gt;22&lt;/sup&gt; /m&lt;sup&gt;3&lt;/sup&gt;. Voids are not observed obviously. The post-irradiation annealing causes the loop size to increase and the loop density to decrease, with numbers of (18.25±16.92) nm and 1.19×10&lt;sup&gt;22&lt;/sup&gt; /m&lt;sup&gt;3&lt;/sup&gt;, respectively. Through the contrast-extinction analysis, the dislocation loops in the irradiated and annealed specimen are identified to be &lt;i&gt; &lt;b&gt;a&lt;/b&gt; &lt;/i&gt;/2&lt;inline-formula&gt;&lt;tex-math id="Z-20230218170937"&gt;\begin{document}$\langle {111} \rangle$\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="5-20222124_Z-20230218170937.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="5-20222124_Z-20230218170937.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;-type dislocation loops. The post-irradiation annealing also causes the coalescence of large loops and forms large irregular-shaped dislocation loops. Voids with typical sizes of 1–2 nm are also observed in the annealed specimen. The PIA modified microstructure is consistent with the stage IV or stage V characterization of classical PIA induced microstructures. Through the “one-step” inside-outside contrast method, the dislocation loops are identified as an interstitial type. The influcing factors for the “one-step” method are discussed and the importance of consistent indexing is also noted.

Simulation of neutron irradiation-induced recrystallization of tungsten

&lt;sec&gt;Tungsten is the candidate for divertor target material in future fusion reactors. The tungsten divertor target is expected to long serve in a harsh environment of high temperature and high-energy neutron irradiation. This can lead to neutron irradiation-induced recrystallization of tungsten, thereby increasing the possibility of intergranular brittle failure and compromising the safe operation of the divertor. Thus, clarifying the mechanism of neutron irradiation-induced tungsten recrystallization is important. However, the current model, which only considers the irradiation-enhanced effect on recrystallization driving force, underestimates the irradiation effect on recrystallization compared with the results observed in recent high-temperature neutron irradiation experiments in the HFIR reactor. It indicates that other irradiation effects can also influence the recrystallization process.&lt;/sec&gt;&lt;sec&gt;In this study, we introduce the irradiation-enhanced grain boundary migration factor (&lt;i&gt;R&lt;/i&gt;) into the established irradiation-induced recrystallization kinetic model, on the assumption that the grain boundary migration velocity is proportional to the self-diffusion coefficient. The simulation results show that after considering both irradiation-enhanced recrystallization driving force and grain boundary migration effect, the calculated half-recrystallization time (&lt;inline-formula&gt;&lt;tex-math id="M4"&gt;\begin{document}${t}_{{X}\text{}=\text{}0.5}$\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;) at 850 ℃ from the model matches the one obtained in the neutron irradiation experiment in the HFIR reactor. This result indicates that the irradiation-enhanced grain boundary migration effect is one of the important factors affecting irradiation-induced recrystallization. In addition, the difference between irradiated and unirradiated &lt;i&gt;t&lt;/i&gt;&lt;sub&gt;&lt;i&gt;X&lt;/i&gt;=0.5&lt;/sub&gt; decreases with temperature increasing. This phenomenon is due to the fact that as the temperature increases, the contribution of irradiation defects to the driving force for recrystallization decreases owing to the irradiation defect recombination. Moreover, the increase of thermal activation diffusion coefficient is more significant than the increase of the irradiation-enhanced diffusion coefficient. These findings suggest that the thermal activation effect eventually dominates the recrystallization process over the irradiation effect as temperature increases.&lt;/sec&gt;

Future for inertial-fusion energy in Europe: a roadmap

Abstract The recent achievement of fusion ignition with laser-driven technologies at the National Ignition Facility sets a historic accomplishment in fusion energy research. This accomplishment paves the way for using laser inertial fusion as a viable approach for future energy production. Europe has a unique opportunity to empower research in this field internationally, and the scientific community is eager to engage in this journey. We propose establishing a European programme on inertial-fusion energy with the mission to demonstrate laser-driven ignition in the direct-drive scheme and to develop pathway technologies for the commercial fusion reactor. The proposed roadmap is based on four complementary axes: (i) the physics of laser–plasma interaction and burning plasmas; (ii) high-energy high repetition rate laser technology; (iii) fusion reactor technology and materials; and (iv) reinforcement of the laser fusion community by international education and training programmes. We foresee collaboration with universities, research centres and industry and establishing joint activities with the private sector involved in laser fusion. This project aims to stimulate a broad range of high-profile industrial developments in laser, plasma and radiation technologies along with the expected high-level socio-economic impact.

Simulation study on uniaxial tensile mechanical and thermal properties of rhenium tungsten alloys

As an important potential first wall material in fusion reactor, tungsten will be irradiated by 14.1 MeV fusion neutron and then transmutated to rhenium, osmium, and tantalum during service. Due to rhenium is the main of transmutation products, which will affect the mechanical and thermal properties of the tungsten matrix. Based on this, in this work, the Young's Modulus and thermal properties of W-Re alloys with different rhenium contents were calculated by using the first principle calculations. Then, the uniaxial tensile processes in [100], [110], [111] and [211] directions were simulated by molecular dynamic simulation method. The results show that the tensile strength of W-Re alloy decreases gradually with the increase of rhenium content. With the increase of the rhenium content, the heat capacity of the material will decrease, and the thermal expansion coefficient increases.

Surface-limited deuterium uptake of Ru films under plasma exposure

Blister formation has been an emerging research topic for extreme ultraviolet (EUV) mirrors exposed to hydrogen plasmas. Similar to plasma-facing materials in nuclear fusion reactors, it has been reported that blister formation in EUV mirrors is initiated by hydrogen uptake due to hydrogen ion or atom bombardment. However, the research so far has focused on Mo/Si multilayers exposed to only hydrogen ions or atoms, while the EUV mirror typically has a Ru capping layer facing hydrogen plasmas. We present experimental work to measure plasma-induced hydrogen uptake of Ru films. We bombarded our designed Ru-capped target with a low-temperature deuterium plasma and measured the deuterium retention using elastic recoil detection. Contrary to ion-driven deuterium uptake, the deuterium uptake rate of the Ru film had no dependence on the deuterium ion flux or energy after a period of plasma exposure. A reaction–diffusion model has been built to calculate the time evolution of deuterium retention, which well fits the experimental data. Based on this model, we conclude that the surface composition of the Ru film is the limiting factor for the deuterium uptake, which is seriously weakened when the surface is covered by Ru oxide. After the Ru oxide is reduced by the plasma, the uptake rate is predominantly driven by the deuterium surface coverage on metallic Ru. Our model also indicates that at the deuterium-populated Ru surface, deuterium has a low absorption barrier to penetrate the surface, which is supported by previously reported computational work.

Application of the Radioluminography Method for the Analysis of the Distribution of Small Amounts of Tritium in Electron-Induced Defects in Advanced Materials for Fusion Reactors

Application of the Radioluminography Method for the Analysis of the Distribution of Small Amounts of Tritium in Electron-Induced Defects in Advanced Materials for Fusion Reactors

Accurate Fe–He machine learning potential for studying He effects in BCC-Fe

Nanostructured Ferritic Alloys (NFAs) containing oxide nanoparticles are prospective advanced structural materials in future fusion reactors. Lacking fusion neutron sources with an adequate flux, modeling is critical to understand radiation damage in the NFAs, including helium effects. Machine learning interatomic potentials (MLPs) have been demonstrated to be capable of describing atomic interactions in chemically complex systems with comparable accuracy to density functional theory (DFT) but with much less computational costs. Hence, MLPs are chosen to study He bubble accumulation in chemically complex NFAs. As a preliminary step, we develop a Fe–He potential, based on ∼10,000 atomic configurations. The developed MLP accurately predicts the bulk properties of BCC-Fe compared to DFT, including the lattice constant (2.832 Å), elastic constants (c11=271 GPa, c12=141, c44=93), and phonon frequencies (maximum error < 3.5%). The MLP predicts He binding energies in HenV bubbles and Hen clusters with a mean absolute error (MAE) of only ∼70 meV, which is ∼3 to ∼5 times smaller than the MAE of binding energies estimated using empirical potentials. The MLP also accurately predicts the migration barrier of interstitial He (64 meV). Furthermore, the MLP correctly predicts the maximum number of He atoms (critical size) at which a HenV bubble (Hen cluster) can kick-out a self-interstitial atom (SIA), forming a bigger HenV2 bubble (HenV bubble). Subsequently, we use the MLP to perform molecular dynamics simulations to investigate the effect of temperature on the critical size of HenV bubbles and Hen clusters. We find that the critical size decreases with increasing temperature — indicating that smaller bubbles and clusters can kick-out an SIA at higher temperatures, thereby escalating the He bubble growth. The current Fe–He MLP can be further developed to include all chemical interactions in NFAs.