Fusion Reactor Components Research Articles

Effect of W-Cu joining on D transport behavior in plasma-facing components for fusion reactors

In future fusion reactors, W-Cu joining technology will be utilized to fabricate plasma facing components (PFCs). In this study, hydrogen isotope gas-driven permeation (GDP) and thermal desorption spectrometry (TDS) are performed for W-Cu, W and Cu samples to understand the impact of W-Cu joining on fuel transport. GDP results reveal that the diffusion activation energy of the W-Cu sample is 0.289 eV, significantly smaller than that of the reference sample. Compared with the reference sample, the deuterium (D) retention in the W-Cu sample is doubled. High resolution transmission electron microscope (HRTEM) images indicate the presence of a mixed region of about 20 nm at the W-Cu interface, with a dislocation density in this region of 7.38 × 1016 m−2, three orders of magnitude higher than that in W and Cu. Simultaneously, energy dispersive spectrometer (EDS) data show that significant oxygen (O) impurities within the mixed region. Results from first principles simulations suggest that the presence of O at the interface can significantly alter the formation energy of H. The mixed layer not only serves as a diffusion barrier but also has the capacity to capture D. The presence of a high number of dislocations and O impurities in the mixed layer greatly impacts D transport in the W-Cu component.

Swift heavy ion irradiation effects on tungsten carbide films

Tungsten carbide (WC) is an important material to study for various applications in extreme environments due to its radiation-resistant properties and high mechanical strength. Due to its excellent properties such as high hardness, strength, and wear resistance, it is also being considered lately as a plasma facing component in nuclear fusion reactors. To study the response of this material in extreme radiation environments, WC films prepared using RF sputtering technique were irradiated with 100 MeV Ag8+ ions in the present work. The irradiation was carried out at three different fluences: 3 × 1012, 1 × 1013, and 3 × 1013 ions/cm2. Rutherford Backscattering Spectrometry (RBS) was performed to determine the stoichiometry and thickness of the pristine film. To study the irradiation-induced modifications, glancing angle X-ray diffraction (GAXRD), field emission scanning electron microscopy (FE-SEM), High Resolution Transmission Electron Microscopy (HRTEM), and Raman spectroscopy were performed on the pristine and irradiated samples. GAXRD of the pristine and irradiated samples revealed grain size reduction accompanied with reduced crystallinity with increasing ion fluence which was further manifested in HRTEM measurements and Raman spectroscopy measurements. Change in surface morphology due to the ion irradiation was observed in FE-SEM. These modifications have been explained using thermal-spike and TRIM simulations which showed that although the swift heavy ions irradiation did not result in track formation in the WC films, it caused radiation damage which may be attributed to the synergistic effect of electronic and nuclear energy loss mechanisms. These results might be crucial for a fundamental understanding of the radiation resistance of WC thin films and its application in different radiation environments.

Thermo-hydraulic performance of tungsten flat tile type divertor target with two types of novel engineered cooling channel wall

The divertor is one of the key components in magnetic confinement fusion reactors. Improving the heat transfer performance of the divertor has great importance for promoting its service capacity. This work proposed a new kind of cooling surface configuration for a flat-type divertor design, which is based on the concept of synergetic heat transfer enhancement via combination of a sinusoidal undulation structure with a micro-nano structured surface (copper-mesh brazed coating). The combination surface configuration is preliminarily obtained by numerical optimization based on the calibrated model of subcooled flow boiling heat transfer. Then, experimental mockups were fabricated and tests with high heat flux (HHF) conditions were conducted to examine the effects of the new design. Both the numerical and experimental results confirm that the incident critical heat flux is increased from 15 MW/m2 of the micro-nano structured flat surface case to over 22 MW/m2 of the combination case under the conditions of relatively low flow velocity of 3 m/s and low flow resistance increase of 7.3 %.

Initial experiments to regenerate the surface of plasma-facing components by wire-based laser metal deposition

Plasma-facing components (PFC) in nuclear fusion reactors are exposed to demanding conditions during operation. The combination of thermal loads, plasma exposure as well as neutron induced damage and activation limits the number of materials suitable for this application. Due to its properties, tungsten (W) is foreseen as plasma-facing material (PFM) for the future DEMOnstration power plant. It is considered suitable due to its exceptionally high melting point, excellent thermal conductivity, low tritium retention and low erosion resistance during plasma exposure. But even tungsten armored PFCs have a limited lifetime due to, among other factors, surface erosion and the resulting thickness reduction of the armor material.In-situ local deposition of tungsten by means of additive manufacturing (AM) could counteract surface erosion and thus increase the service life span of PFCs. After evaluation of the potential AM processes qualified for this task, the wire-based laser metal deposition (LMD-w) process was selected as the most suitable process. First trials were conducted to examine if it is possible to reliably deposit tungsten onto tungsten substrate using the LMD-w process. In these first studies, single welding beads were generated, and in later experiments, entire layers were created from several welding beads which are arranged next to each other. To ensure reproducibility of the results, the substrate temperature was kept constant. Further experiments aimed at the elimination or minimization of problems such as oxidation, occurrence of balling defects, porosity, cracking, surface waviness and insufficient connection to the substrate. To increase the welding bead quality, the input parameters like laser power, deposition velocity, wire feed rate, inert gas flow, as well as the wire position were optimized. Furthermore, stacking of several layers, as well as the remelting of an already created layer, were carried out and investigated. This study represents the first steps in testing the feasibility of an in-situ surface regeneration concept for PFCs.

Prospects for melt refining of radioactive waste from fusion reactors

Fusion has the prospect of producing only low-level radioactive waste, but the volume of this waste is potentially very large. A possible means of reducing the volume of this waste is to recycle the activated materials, e.g. within the fusion industry. In the process of steelmaking, certain impurities are removed via accumulation in slag or fumes produced during the process, and this has been successfully employed as a means of clearing surface-contaminated LWR steels (e.g. steam generators). In this work, we simulate the process applied to neutron-activated fusion reactor components including the blanket, divertors, and shielding.The process is based on a well known partitioning between melt, slag, dust, and gas that is based on the oxidation potential of the elements and their volatility. We model this partitioning of activated materials and use FISPACT-II to simulate their activation during operation; the processes are repeated cyclically to simulate the long-term accumulation of activation products in the steel and waste products. The principal findings are that most of the induced activity remains in the metal rather than accumulating in the slag, and that while the slag remains a low-level waste under the assumptions employed, its volume actually exceeds that of the recycled material, and it therefore did not constitute a successful waste volume reduction technique.

Predictive atmospheric dispersion and deposition characteristics of activated tungsten dust

Tungsten is considered to be the most promising material for plasma-facing components of fusion reactors in addition to other industrial and military uses. Activated tungsten dust produced by plasma-wall interaction and released under accidental conditions would be the emerging environmental pollutant with both chemical and radiation toxicity. For the first time, atmospheric dispersion and deposition characteristics were investigated for tungsten dust with various particle size spectrum similar as generated in fusion reactors. Euler-Lagrange approach was adopted to describe parcels motion and deposition behaviors. Proper numbers of parcels were determined for the atmospheric dispersion and various parcel diameters scale. Near-surface concentration distribution and dry deposition velocity were predicted for tungsten dust with both single size dust and realistic particle size spectrum dust. Influences of chemical forms and release factors on atmospheric dispersion and deposition of tungsten dust were preliminary identified at last. It is indicated that 150 million parcels are needed for 10 μm dust under 5 km scale. Dry deposition velocity of tungsten dust is up to 11.3 cm s−1 and increase along with particle diameter because of its high density and gravitational deposition effect. The critical diameter judging whether gravitational deposition could appear for tungsten dust is 1 µm. Deposited particle size spectrum would be completely different along the downwind distance which would result in differences in subsequent behaviors in other environmental compartments.

Study of irradiation-induced vacancy-type defects in tungsten using positron annihilation spectroscopy

Tungsten has been considered as a potential candidate material for the application of plasma-facing components in magnetic confinement fusion reactors. Fusion neutrons introduce severe displacement damage into the first-wall materials along with a high amount of helium which is detrimental to the desirable material properties. Ion beam implantation combined with positron annihilation spectroscopy is a well-adopted method to investigate the irradiation damage in materials without any induced radioactivity. In this study, two separate tungsten samples were individually irradiated with 9.1 MeV Au5+ ions to a dose of 2.5 × 1014 ions/cm2 and 130 keV He ions to a dose of 1 × 1016 ions/cm2 with the intention to investigate the evolution of vacancy-type defects in the absence and presence of helium. The irradiated samples were subjected to isochronal annealing and characterized using positron annihilation Doppler broadening spectroscopy. The evolution of vacancy-type defects during different stages of annealing was investigated using defect-sensitive positron S-parameter. An increase in S-parameter due to the presence of vacancy-type defects was observed upon irradiation in both samples. The S-parameter variations with respect to the temperature showed different stages of vacancy annealing corresponding to stage III from 473 K to 673 K where the migration and clustering of mono-vacancies occurs, stage IV from 673 K to 973 K due to the growth of nano-voids and stage V from 973 K to 1273 K due to the dissociation of nano-voids. Complete recovery of irradiation-induced defects occurred in both samples at 1273 K. Both the samples showed similar vacancy annealing stages and no signature on the formation and growth of He-vacancy clusters or He bubbles was observed in the present study.

Numerical and experimental development of cupronickel filler brazed joints for divertor and first wall components in DEMO fusion reactor

The brazeability of a cupronickel commercial alloy (Cu10Ni) was evaluated for its use as a filler alloy for high-temperature joining of tungsten to the reduced activation ferritic/martensitic steel EUROFER 97 (W-E) and between tungsten base materials (W-W) for its application at the first wall and divertor of future fusion reactors. In addition, given the importance of the residual stresses in these heterogenous joints, a study of the brazing conditions and the impact of the selected filler has been conducted using numerical software to understand its impact on the quality of the joint.Two thermal cycles were evaluated (1165 °C and 1190 °C) and selected based on the thermal characterization of the filler alloy. The microstructural examination revealed that, in W-E joints, nickel acts as an activator element, reacting and forming interfacial layers at the EUROFER 97 - Cu10Ni interface. In the case of the W-W joints, a lower level of diffusion phenomenon and metallurgical interaction between Cu10Ni and base materials were observed. The hardness profile indicated that the hardening process of EUROFER 97 was associated with the formation of untempered martensite. On the other hand, tungsten kept the received hardness. The mechanical characterization by shear test reported similar values between both types of joints carried out at 1190 °C but different when the temperature was increased (1165 °C), associated with the brittle character of tungsten and its lower metallurgical interaction.The numerical analysis of the brazing process carried out with ANSYS software shows that residual stresses are accumulated mainly at the interfaces. The information provided by the simulation shows, for a 50 µm filler thickness, the importance of mitigating the residual stress by selecting a filler with an intermediate Coefficient of Thermal Expansion (CTE) that alleviates mechanical stresses relative to the base materials.

Microstructure and Magnetism of Heavily Helium-Ion Irradiated Epitaxial Iron Films

This study reports on the microstructure and magnetism of pure iron irradiated with high doses of helium ions. Iron alloys are important structural materials used as components in fusion reactors, and a comprehensive database of their various properties has been developed. But little has been investigated on magnetic properties, in particular, the effects of high doses and helium cavities are lacking. Single-crystal iron films, with a thickness of 200 nm, were prepared using the ultra-high vacuum evaporation method. These films were then irradiated with 30 keV He+ ions at room temperature up to a dose of 18 dpa. X-ray diffraction measurements and cross-sectional transmission electron microscope observations revealed significant microstructural changes, including a large lattice expansion perpendicular to the film plane and the formation of high-density cavities after irradiation. However, the saturation magnetization and the shape of the magnetization curve showed almost no change, indicating the robustness of the magnetic properties of iron.

Recrystallization kinetics of larger yttria particle dispersion tungsten alloy warm rolled to 50% thickness reduction

Tungsten-based materials are considered one of the candidate materials for the key components (such as the first wall and diverter) of future thermonuclear fusion reactors because of their excellent properties. At high operation temperatures, restoration processes such as recovery, recrystallization and grain growth will unavoidably embrittle ductile tungsten achieved by rolling. Dispersed second-phase particles will affect the recrystallization behavior. The thermal stability of a yttria dispersion-strengthened tungsten plate warm-rolled to 50% thickness reduction is investigated. Isochronous annealing is used to predict its recrystallization temperature range and isothermal annealing is used to study its recrystallization kinetics. The tungsten alloy in the present study has a faster recrystallization kinetics and a smaller activation energy compared to that of W-2 vol% Y2O3 alloy with small Y2O3 particles, indicating that the larger Y2O3 particles have little effect on impeding grain boundaries during recrystallization.

Effects of grain boundaries on the evolution of radiation-induced bubbles in polycrystalline tungsten: A phase-field simulation

Tungsten (W) is widely recognized as a promising material for plasma-facing components in nuclear fusion reactors, and the evolution of bubbles in polycrystalline tungsten significantly impacts its properties. This study developed a phase-field model considering elastic inhomogeneity to investigate the evolution of radiation-induced bubbles. The Gibbs free energy of helium bubbles was derived using the Van der Waals equation. An explicit nucleation algorithm accounting for the position of grain boundaries (GBs) was coupled to the model. Furthermore, the model also considered the inequality between the internal pressure of the bubble and the surface tension. The growth kinetics of a single bubble with elastic heterogeneity were examined, and based on these findings, the evolution of bubbles under irradiation was simulated, yielding results consistent with experimental observations. Furthermore, the influence of GBs on radiation-induced bubbles was investigated using the developed model. The simulation results demonstrated that both GBs, as preferential nucleation regions and as sinks for vacancy absorption, had an impact on bubble nucleation and growth. The superior radiation resistance observed in nanocrystalline tungsten was dominantly attributed to the absorption of vacancies by GBs.

Hydrogen isotope permeation through interfaces and permeability of tungsten layers

The estimation of the hydrogen isotope flux through a fusion reactor wall component is important for the material selection and in order to guarantee a safe and economical reactor operation. Since the permeation flux through a component cannot be measured directly, due to the large size of such a component, the deuterium permeability of various fusion materials were investigated in the last years. In order to investigate, if an interface, meaning a change of materials in a component, has an influence on the permeation flux, combined material systems in laboratory scale are studied. The combined materials systems are produced by applying a thin W or Cu layer on a polished bulk steel or CuCrZr substrates by magnetron sputter deposition. In a previous study (Houben et al., 2022), the combined material system of Cu coated steel was investigated. The conclusion was that the interface has a minor influence on the permeation flux compared to the large influence of the layer microstructure. In order to investigate, if the minor influence of the interface is valid in general, the systems W coated steel and CuCrZr are studied in this publication. By heating the substrate during W sputter deposition, a crack-free W layer is produced and crack propagation in the W layer at elevated temperature is prevented. The W layer permeabilities are obtained for both W coated substrates and are similar, but compared to the W bulk permeability from literature the W layer permeabilities are several orders of magnitude larger.The main conclusion from these studies is that in all investigated combined material systems the influence of the interface on the permeation flux is minor compared to the large influence of the microstructure. Therefore, for a reliable estimation of the permeation flux through a fusion reactor component it is crucial to characterize the applied materials. Especial for coatings the measurement of the layer permeability is important, since the layer permeability of a material can be very different compared to the permeability of a bulk material.

Physics design of 14 MeV neutron generator facility at the Institute for Plasma Research

A high energy and high yield neutron source is a prime requirement for technological studies related to fusion reactor development. It provides a high-energy neutron environment for small-scale fusion reactor components research and testing such as tritium breeding, shielding, plasma-facing materials, reaction cross-section data study for fusion materials, etc. Along with ITER participation, the Institute of Plasma Research, India is developing an accelerator-based 14 MeV neutron source with a yield of 1012 n s−1. The design of the source is based on the deuterium–tritium fusion reaction. The deuterium beam is accelerated and delivered to the tritium target to generate 14 MeV neutrons. The deuterium beam energy and tritium availability in the tritium target are the base parameters of the accelerator-based neutron source design. The paper gives the physics design of the neutron generator facility of the Institute for Plasma Research. It covers the requirements, design basis, and physics parameters of the neutron generator. As per the analytical results generator can produce more than 1 × 1012 n s−1 with a 110 keV D+ ion beam of 10 mA and a minimum 5 Ci tritium target. However, the detailed simulation with the more realistic conditions of deuteron ion interaction with the tritium titanium target shows that the desired results cannot be achieved with 110 keV. The safe limit of the ion energy should be 300 keV as per the simulation. At 300 keV ion energy and 20 mA current, it reaches 1.6 × 1012 n s−1. Moreover, it was found that to ensure sufficiently long operation time a tritium target of more than 20 Ci should be used. The scope of the neutron source is not limited to the fusion reactor research studies, it is extended to other areas such as medical radioisotopes research, semiconductor devices irradiations, and many more.

Construction of a new n-body potential and multi-scale investigations of the direct alloying behaviors for immiscible W/Cu system

Achieving direct alloying at the tungsten/copper (W/Cu) interface is crucial for the fabrication of laminar metal matrix composite (LMMC) plasma-facing components (PFCs) of fusion reactors based on immiscible W/Cu systems. Our previous work has accomplished direct alloying between immiscible W and Cu, forming a diffusion layer with a thickness of approximately 22–32 nm at the interface. The performance of the interface directly affects the lifetime and reliability of W/Cu LMMC. Multi-scale calculations are designed since some interface properties cannot be directly measured or observed experimentally in real-time and in situ. Firstly, a new potential is constructed to overcome the limitations that the existing W/Cu potential cannot accurately describe the W/Cu alloying process. Secondly, molecular dynamics (MD) simulations use the new potential to study interfacial properties and calculate the diffusion depth at the W/Cu interface. Lastly, combining MD, FEM, and image visualization techniques, a multi-scale method of the crack propagation rate at the W/Cu interface is used to predict the thermal fatigue life of W/Cu LMMC PFCs prepared by nano-active structure-induced alloying. Our simulation results agree with experimental values and literature results, demonstrating the reliability of the new W/Cu potential and the accuracy of our designed multi-scale method. Directly alloyed W/Cu LMMCs have the potential to be utilized in fusion reactors.

Research and development of copper joint for ITER In-Vessel Coils

The ITER In-Vessel Coils (IVCs) are crucial components of the Tokamak fusion reactor. They are located within the vacuum vessel (VV) behind the blanket shield modules and are responsible for compensating for fast perturbations of the plasma. There are a total of 176 joints used to connect the IVCs, feeders and feedthrough. The joint structure consists of a central copper conductor, a mineral (MgO) insulation layer, and an outer stainless-steel jacket. Given the inaccessibility of the IVCs once the blanket modules are installed, the highest level of quality is required for the joints. One key item is the welding of the copper conductor which has to be done inside the VV. To address the high thermal conductivity of copper conductor and the challenging constraints working inside the VV, a preheating method by the arc of the welding head has been developed. Macro- and micro-structure examination were performed on the welding cross sections, and tensile tests were carried out to check the strength of the copper joint, which demonstrated that arc preheating does not have a significant impact on the heat-affected zone. After more than 40 welding attempts and several design updates, the copper joint has been successfully developed and can be repeatedly and successfully welded.

Review of Recent Progress in Plasma-Facing Material Joints and Composites in the FRONTIER U.S.-Japan Collaboration

The plasma-facing components (PFCs) of future fusion reactors will have intricate structures and require multiple materials because no one material can simultaneously satisfy all the requirements of the component. The dissimilar material joints in PFCs must withstand extreme thermal and stress gradients under neutron irradiation. The Fusion Research Oriented to Neutron Irradiation and Tritium Behavior at Material Interfaces (FRONTIER) U.S.-Japan collaboration seeks to explore and explain the behavior of internal solid interfaces in PFCs under neutron irradiation. The first step of the collaboration was to identify the leading PFCs that should be studied further and prepare them for the next step, which will include neutron irradiation. Different strategies for material development are being pursued worldwide to produce robust PFCs. Here, an overview is presented of some of the most promising materials in the areas of copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites. Each material’s fabrication and benefits are described, and some discussion of remaining questions is given.

Parametric Study of the Vacuum Permeator for the Tritium Extraction eXperiment

Tritium breeding is a critical component of any self-sustaining future fusion reactor. The liquid-metal eutectic PbLi is of particular interest as a tritium breeder material due to its favorable thermophysical and neutronic properties. One of the several remaining challenges facing PbLi breeder blankets is the need to design and validate a highly efficient tritium extraction system. The vacuum permeator is a promising extraction concept that utilizes tritium permeation through a highly permeable metal membrane. The Tritium Extraction eXperiment (TEX) is a forced-convection PbLi loop constructed to investigate tritium extraction from PbLi with vacuum permeators. Accurate thermal-hydraulic and tritium transport models are required to establish appropriate test matrices, predict experiment outcomes, and analyze data. However, the hydrogen transport properties of PbLi and permeator materials have large uncertainties. A database is collected and a parametric analysis is conducted on the effect of hydrogen transport material properties, including diffusivity of H in PbLi and the permeator, solubility of H in PbLi and the permeator, and the permeator surface recombination constant, on the expected tritium extraction efficiency for a vacuum permeator installed in TEX. Herein, we observe that the solubility of H in PbLi and the permeator and the recombination constant of the permeator have the largest effect on extraction efficiency.

Global sensitivity analysis of a coupled multiphysics model to predict surface evolution in fusion plasma–surface interactions

We construct a global sensitivity analysis framework for a coupled multiphysics model used to predict the changes in material properties and surface morphology of helium plasma-facing components in future fusion reactors. The model combines the particle dynamics simulator F-TRIDYN, that predicts the helium implantation profile, with the cluster dynamics simulator Xolotl, that predicts the growth and evolution of subsurface helium gas bubbles. In order to keep the sensitivity analysis tractable, we first construct a sparse, high-dimensional polynomial chaos expansion surrogate model for each output quantity of interest, which allows the efficient extraction of sensitivity information. The sensitivity analysis is performed for two problem settings: one for ITER-like conditions, and one that resembles conditions inside the PISCES-A linear plasma device. We present a systematic comparison of important parameters, for both F-TRIDYN and Xolotl in isolation as well as for the coupled model, and discuss the physical interpretation of these results.

Bulk Tungsten Fiber-Reinforced Tungsten (Wf/W) Composites Using Yarn-Based Textile Preforms

The use of tungsten fiber-reinforced tungsten composites (Wf/W) has been demonstrated to significantly enhance the mechanical properties of tungsten (W) by incorporating W-fibers into the W-matrix. However, prior research has been restricted by the usage of single fiber-based textile fabrics, consisting of 150 µm warp and 50 µm weft filaments, with limited homogeneity, reproducibility, and mechanical properties in bulk structures due to the rigidity of the 150 µm W-fibers. To overcome this limitation, two novel textile preforms were developed utilizing radial braided W-yarns with 7 core and 16 sleeve filaments (R.B. 16 + 7), with a diameter of 25 µm each, as the warp material. In this study, bulk composites of two different fabric types were produced via a layer-by-layer CVD process, utilizing single 50 µm filaments (type 1) and R.B. 16 + 7 yarns (type 2) as weft materials. The produced composites were sectioned into KLST-type specimens based on DIN EN ISO 179-1:2000 using electrical discharge machining (EDM) and subjected to three-point bending tests. Both composites demonstrated enhanced mechanical properties with pseudo-ductile behavior at room temperature and withstood over 10,000 load cycles between 50–90% of their respective maximum load without sample fracture in three-point cyclic loading tests. Furthermore, a novel approach to predict the fatigue behavior of the material under cyclic loading was developed based on the high reproducibility of the composites produced, especially for the composite based on type 1. This approach provides a new benchmark for upscaling endeavors and may enable a better prediction of the service life of the produced components made of Wf/W in the future. In comparison, the composite based on fabric type 1 demonstrated superior results in manufacturing performance and mechanical properties. With a high relative average density (>97%), a high fiber volume fraction (14–17%), and a very homogeneous fiber distribution in the CVD-W matrix, type 1 shows a promising option to be further tested in high heat flux tests and to be potentially used as an alternative to currently used materials for the most stressed components of nuclear fusion reactors or other potential application fields such as concentrated solar power (CSP), aircraft turbines, the steel industry, quantum computing, or welding tools. Type 2 composites have a higher layer spacing compared to type 1, resulting in gaps within the matrix and less homogeneous material properties. While type 2 composites have demonstrated a notable enhancement over 150 µm fiber-based composites, they are not viable for industrial scale-up unlike type 1 composites.

Double-sided friction stir welding of Nitronic-40 stainless steel for application in tokamak devices

Nitrogen containing austenitic stainless steels (N-ASS) are widely utilized to fabricate various structural components in tokamak type fusion reactors owing to their suitable mechanical and functional properties. These components are exposed to a range of temperatures (4–500 K) and interact closely with the magnetic fields that are used to control and contain the plasma within the tokamak systems. Nitronic-40 (N40) or XM-11 stainless steel is one such N-ASS used for fabricating structural components in the magnetic and vacuum vessel systems in tokamak devices. Fabrication of most of the larger components in the magnetic and vacuum vessel systems typically involves some type of fusion-based welding process. This study presents a double-sided friction stir welding (FSW) approach as an alternative to fusion welding processes to join 12 mm thick N40 plates to obtain joints with a low fraction of δ ferrite (a detrimental ferromagnetic phase), high joint efficiency, no sensitization and loss of hardness in the heat affected zone, and minimal nitrogen desorption from the weld nugget. The double-sided FSW approach yielded superior weldments when compared to similar joints accomplished by fusion welding for application in tokamak devices.