Search Results (36)

Copper alloys are critical heat sink materials for fusion reactor divertors due to their high thermal conductivity (TC) and strength, yet their performance under extreme particle bombardment and heat fluxes in future tokamaks requires enhancement. While neutron-induced transmutation helium affects the properties of copper, the atomistic mechanisms linking helium bubble size to thermal transport remain unclear. This study employs non-equilibrium molecular dynamics (NEMD) simulations to isolate the effect of bubble diameter (10, 20, 30, 40 Å) on TC in copper, maintaining a constant He-to-vacancy ratio of 2.5. Results demonstrate that larger bubbles significantly impair TC. This reduction correlates with increased Kapitza thermal resistance and pronounced lattice distortion from outward helium diffusion, intensifying phonon scattering. Phonon density of states (PDOS) analysis reveals diminished low-frequency peaks and an elevated high-frequency peak for bubbles >30 Å, confirming phonon confinement and localized vibrational modes. The PDOS overlap factor decreases with bubble size, directly linking microstructural evolution to thermal resistance. These findings elucidate the size-dependent mechanisms of helium bubble impacts on thermal transport in copper divertor materials. Full article

The Volumetric Neutron Source (VNS) is a pivotal facility proposed for advancing fusion nuclear technology, particularly for the qualification of breeding blanket systems, a key component of DEMO and future fusion reactors. This study focuses on the design and optimisation of the VNS Thermal Shield, adopting a multidisciplinary approach to address its thermal and structural behaviours. The Thermal Shield plays a crucial role in protecting superconducting magnets and other cryogenic components by limiting heat transfer from higher-temperature regions of the tokamak to the cryostat, which operates at temperatures between 4 K and 20 K. To ensure both thermal insulation and structural integrity, multiple design iterations were conducted. These iterations aimed to reduce electromagnetic (EM) forces induced during magnet charge and discharge cycles by introducing strategic cuts and reinforcements in the shield design. The optimisation process included the evaluation of various aluminium alloys and composite materials to achieve a balance between rigidity and weight while maintaining structural integrity under EM and mechanical loads. Additionally, an integrated thermal study was performed to ensure effective temperature management, maintaining the shield at an operational temperature of around 80 K. Cooling channels were incorporated to homogenise temperature distribution, improving thermal stability and reducing thermal gradients. This comprehensive approach demonstrates the viability of advanced material solutions and design strategies for thermal and structural optimisation. The findings reinforce the importance of the VNS as a dedicated platform for testing and validating critical fusion technologies under operationally relevant conditions. Full article

The exploitation of fusion energy in tokamak reactors relies on efficient and reliable tritium management. The tritium needed to sustain the deuterium–tritium fusion reaction is produced in the Li-based blanket surrounding the plasma chamber, and, therefore, the effective extraction and purification of the tritium bred in the Li-blankets is needed to guarantee the tritium self-sufficiency of future fusion plants. This work introduces a new technology for the extraction of tritium from the Pb–Li eutectic alloy used in liquid blankets. Process units based on the concept of Membrane Gas–Liquid Contactor (MGLC) have been studied for the extraction of tritium from the Pb–Li in the Water Cooled Lithium Lead blankets of the DEMO reactor. MGLC units have been preliminarily designed and then compared in terms of the permeation areas and sizes with the tritium extraction technologies presently under study, namely the Permeator Against Vacuum (PAV) and the Gas–Liquid Contactors (GLCs). The results of this study show that the DEMO WCLL tritium extraction systems using MGLC require smaller permeation areas and quicker permeation kinetics than those based on PAV (Permeator Against Vacuum) devices. Accordingly, the MGLC extraction unit exhibits volumes smaller than those of both PAV and GLC. Full article

The paper presents a conceptual study of a neutron source based on a spherical tokamak (ST). The plasma scenario chosen for the ST is non-thermal fusion (hot ion mode), which is extensively used on machines like JET and TFTR deuterium–tritium (DT) experiments, which seems suited for low fusion gain reactors. As demonstrated in experiments, this scenario is a robust tool for neutron production. Starting from a new scaling law of energy confinement tested, approximately, on ST40 spherical tokamak, the parameters of a 15 MW ST DT fusion reactor (ST180) are derived, and a preliminary radial build of the machine is established. Full article

For the large-scale fusion magnets of the International Thermonuclear Experimental Reactor (ITER) tokamak, wound with cable-in-conduit conductors, the application of sophisticated numerical models able to analyse the thermal–hydraulic behaviour during plasma scenarios is of paramount importance to guarantee an adequate stability margin during operating conditions. The SuperMagnet code has been developed by CryoSoft with the intent to simultaneously simulate the electrical, thermal and hydraulic phenomena occurring during the operation of superconducting coils. In this work, the SuperMagnet code is applied to analyse the thermal–hydraulic behaviour of the central solenoid of the ITER tokamak under the plasma scenario. The central solenoid (CS) is composed of six modules for a total amount of 240 pancakes. The software is able to tackle the complex structure of the CS and its cryogenic closed loop. In the present work, the circulation pump operation and the heat transfer to the helium bath are investigated. The results presented here show the temperature evolution of the magnet and of the supercritical helium during the plasma scenario, which allows the determination of the operation margin of the CS. Full article

Nuclear fusion is a promising energy source. The International Thermonuclear Experimental Reactor aims to study the feasibility of tokamak-type reactors and test technologies and materials for commercial use. One major challenge is developing materials for the reactor’s divertor, which supports high thermal flux. Tungsten was chosen as the plasma-facing material, while a CuCrZr alloy will be used in the cooling pipes. However, the gradient between the working temperatures of these materials requires the use of a thermal barrier interlayer between them. To this end, refractory high-entropy (CrFeTiTa)₇₀W₃₀ and VFeTiTaW alloys were prepared by mechanical alloying and sintering, and their thermal and irradiation resistance was evaluated. Both alloys showed phase growth after annealing at 1100 °C for 8 days, being more pronounced for higher temperatures (1300 °C and 1500 °C). The VFeTiTaW alloy presented greater phase growth, suggesting lower microstructural stability, however, no new phases were formed. Both (as-sintered) alloys were irradiated with Ar⁺ (150 keV) with a fluence of 2.4 × 10²⁰ at/m², as well as He⁺ (10 keV) and D⁺ (5 keV) both with a fluence of 5 × 10²¹ at/m². The morphology of the surface of both samples was analyzed before and after irradiation showing no severe morphologic changes, indicating high irradiation resistance. Additionally, the VFeTiTaW alloy presented a lower deuterium retention (8.58%) when compared to (CrFeTiTa)₇₀W₃₀ alloy (14.41%). Full article

In recent years, fusion energy has assumed an important role in the energy scenario, being a sustainable, environmentally friendly, and practically inexhaustible energy source. Fusion energy could play a crucial role in fully decarbonized electricity production in the second half of this century, helping to meet the increasing energy demand. One of the studied reactors is ARC, a tokamak fusion device characterized by a compact and high-field design initially conceived by researchers at the Massachusetts Institute of Technology, which the Commonwealth Fusion System (CFS) plans to construct in the next decade. This paper is focused on the analysis and development of different configurations for the ARC Balance of Plant Power Conversion System, with the aim of improving the thermodynamic efficiency, which is one of the pillars of sustainability. Three cycles were studied by using the General Electric GateCycle^TM software: a supercritical steam Rankine cycle, a supercritical CO₂ Brayton cycle, and a supercritical helium Brayton cycle. The thermal efficiency of the three options was compared to select the most promising solution. The results showed that the supercritical steam cycle is the best configuration in terms of cycle efficiency for the ARC FNSF Pilot phase. Full article

Nuclear fusion is understood as an energy reaction that does not emit greenhouse gases, and it has been considered as a long-term source of low-carbon electricity that is favourable to curtail rapid climate change. Fusion offers a pathway to resolve energy security and the unequal distribution of energy resources since seawater is its ultimate fuel source and a few grams of fuel can generate mega kilowatts of power. The development and testing of new materials and technologies are unceasing to achieve the net fusion energy through national and international collaboration as well as private partnerships. The ever-growing number of research works report various designs and magnet-based fusion devices, such as stellarators, lasers, and tokamaks. This article provides an overview on the utilization of nuclear energy as a clean energy source, as well as the strategies and progress towards establishing successful commercial fusion energy to the grid and transition to a reliable clean energy source. The overview focuses on the fusion nuclear development in five major countries, UK, US, China, Japan, and Russia. Identified technical and financial challenges are also described at the end of this article. The International Thermonuclear Experimental Reactor (ITER) has been an international reference program for fusion energy development and most developed countries with nuclear development capacity are aiming to complete their in-house fusion energy facilities in parallel to ITER. Many fusion programs are finishing the conceptual design and shifting into the phase of engineering design for the planned DEMO fusion facilities. The significant challenges were identified from the perspective of device efficiency and robustness, sustainable funding, and facility maintenance and safety, which must be addressed diligently to realize fusion energy as alternative clean energy that mitigates climate change and supports the goals of energy security. Full article

Nuclear fusion reactors are composed of several complex components whose behavior may be not certain a priori. This uncertainty may have a significant impact on the evolution of fault transients in the machine, causing unexpected damage to its components. For this reason, a suitable method for the uncertainty propagation during those transients is required. The Monte Carlo method would be the reference option, but it is, in most of the cases, not applicable due to the large number of required, repeated simulations. In this context, the Polynomial Chaos Expansion has been considered as a valuable alternative. It allows us to create a surrogate model of the original one in terms of orthogonal polynomials. Then, the uncertainty quantification is performed repeatedly, relying on this much simpler and faster model. Using the fast current discharge in the Divertor Tokamak Test Toroidal Field (DTT TF) coils as a reference scenario, the following method has been applied: the uncertainty on the parameters of the Fast Discharge Unit (FDU) varistor disks is propagated to the simulated electrical and electromagnetic relevant effects. Eventually, two worst-case scenarios are analyzed from a thermal–hydraulic point of view with the 4C code, simulating a fast current discharge as a consequence of a coil quench. It has been demonstrated that the uncertainty on the inputs (varistor parameters) strongly propagates, leading to a wide range of possible scenarios in the case of accidental transients. This result underlines the necessity of taking into account and propagating all possible uncertainties in the design of a fusion reactor according to the Best Estimate Plus Uncertainty approach. The uncertainty propagation from input data to electrical, electromagnetic, and thermal hydraulic results, using surrogate models, is the first of its kind in the field of the modeling of superconducting magnets for nuclear fusion applications. Full article

Fusion reactors are designed to operate at extremely high temperatures, which causes the plasma-facing materials to be heated to 500 °C to 1000 °C. Tungsten is one of the target design materials for the plasma-facing diverter components in Tokamak designs, such as ITER, because of its excellent high-temperature strength and creep properties. However, recrystallization due to high temperatures may be detrimental to these superior mechanical properties, while exposure to He plasma has been reported to influence the recrystallization behaviour. This influence is most likely due to the Zener effect caused by He bubbles formed near the surface, which retard the migration of grain boundaries, while at the same time modifying the surface microstructure. This paper reports a study of the effect of plasma exposure at different sample temperatures on the recrystallization behaviour of W at different annealing temperatures. The characterization after plasma exposure and annealing is pursued through a series of post-exposure annealing, followed by scanning electron microscopy (SEM), electron backscatter diffraction (EBSD) characterization and nanoindentation to determine the mechanical properties. Here, it is shown that the hardness is closely related to the recrystallization fraction, and that the plasma exposure at a sample temperature of 300 °C slows down the recrystallization more than at higher sample temperatures of 500 °C and 800 °C. Atomic force microscopy (AFM) was subsequently used to determine any changes in pile-up height around the nanoindents, to probe any indication of changes in hardenability. However, these measurements failed to provide any clear evidence regarding this aspect of mechanical behaviour. Full article

The construction and operation of the first generation of magnetically controlled nuclear fusion power plants require the development of proper physics and the engineering bases. The analysis of data, recently collected by the actual largest and most important tokamak in the world JET, that has successfully completed his second deuterium and tritium campaign in 2021 (DTE2) with a full ITER like wall main chamber, has provided an important consolidation of the ITER physics basis. Thermonuclear plasmas are highly nonlinear systems characterized by the need of numerous diagnostics to measure physical quantities to guide, through proper control schemes, external actuators. Both modelling and machine learning approaches are required to maximize the physical understanding of plasma dynamics and at the same time, engineering challenges have to be faced. Fusion experiments are indeed extremely hostile environments for plasma facing materials (PFM) and plasma-facing components (PFC), both in terms of neutron, thermal loads and mechanical stresses that the components have to face during either steady operation or off-normal events. Efforts are therefore spent by the community to reach the ultimate goal ahead: turning on the first nuclear fusion power plant, DEMO, by 2050. This editorial is dedicated at reviewing some aspects touched in recent studies developed in this dynamic, challenging project, collected by the special issue titled “New Challenges in Nuclear Fusion Reactors: From Data Analysis to Materials and Manufacturing”. Full article

The in-vessel Loss-of-Coolant Accident (LOCA) is one of the design basis accidents in the design of the EU DEMO tokamak fusion reactor. System-level codes are typically employed to analyse the evolution of these transients. However, being based on a lumped approach, they are unable to quantify localised quantities of interest, such as local pressure peaks on the vacuum vessel walls, to which the failure criteria are linked. To calculate local quantities, the 3D nature of the phenomenon needs to be considered. In this work, a 3D transient model of the in-vessel LOCA from a water-cooled blanket is developed. The model is implemented in the commercial CFD software STAR-CCM+. It simulates the propagation of the water jet in the vessel from the beginning of the accident, thus accounting for the phase change of the water, i.e., from the pressurised liquid phase to the vapour phase inside the vessel, being the latter at a much lower pressure than in the blanket coolant pipes. Due to the large pressure ratio (>1000), shocks are expected; therefore, an Adaptive Mesh Refinement (AMR) algorithm is employed. The physical models (in particular, the multiphase model) are benchmarked to a 2D reference problem before being applied to the 3D EU DEMO-relevant problem. The simulation results show that the pressure peaks in front of the vessel walls are not dangerous as they are below the design limit. The entire evolution of the water jet is followed up to the opening of the burst disks, in order to compare the average pressure evolution with that computed with system-level codes. A comparison with the in-vessel LOCA from a helium-cooled blanket is also carried out, showing that the accident evolution in the water case is less violent than in the helium case. Full article

The China Fusion Engineering Test Reactor (CFETR), a superconducting magnetic confinement tokamak fusion reactor, will develop a high-performance torus cryopump to pump torus plasma exhaust gas. The inlet valve is one of the key components of the cryopump, and it is used to isolate the cryopump from the plasma for regeneration, to control the pumping speed of the cryopump, and to operate as a pressure relief valve in case of a failure, such as the cryopipe breaking inside the cryopump chamber. This paper presents a novel inlet valve. Ensuring that the design of the inlet valve meets the above requirements will be a challenge. In order to verify the reliability of the inlet valve, its critical components are analyzed and optimized by the Finite Element Method. The effect of the stroke of the inlet valve on pumping performance is then estimated by the Monte Carlo Method, and the pressure profile in the whole flow field is studied to predict the cryopump’s behavior. Finally, the seismic capacity of the optimized inlet valve is analyzed, and the mechanical performance of the inlet valve is shown to meet CFETR design criteria. These design and analysis results will provide technical support and references for the development of the CFETR torus cryopump. Full article

Studies on the erosion products of tungsten plasma-facing components (films, surfaces, and dust) for thermonuclear fusion reactors by spectroscopic methods are considered and compared with those of carbon deposits. The latter includes: carbon–deuterium CD_x (x ~ 0.5) smooth films deposited at the vacuum chamber during the erosion of the graphite limiters in the T-10 tokamak and mixed CH_x-Me films (Me = W, Fe, etc.) formed by irradiating a tungsten target with an intense H-plasma flux in a QSPA-T plasma accelerator. It is shown that the formerly developed technique for studying CD_x films with 15 methods, including spectroscopic methods, such as XPS, TDS, EPR, Raman, and FT-IR, is universal and can be supplemented by a number of new methods for tungsten materials, including in situ analysis of the MAPP type using XPS, SEM, TEM, and probe methods, and nuclear reaction method. In addition, the analysis of the fractality of the CD_x films using SAXS + WAXS is compared with the analysis of the fractal structures formed on tungsten and carbon surfaces under the action of high-intensity plasma fluxes. A comparative analysis of spectroscopic studies on carbon and tungsten deposits makes it possible to identify the problems of the safe operation of thermonuclear fusion reactors. Full article

We discuss the general concepts underlying the design strategies of complex devices like Tokamaks and Free Electron Lasers (FEL). Regarding the FEL, starting from the desired output performances, the key parameters are embedded to get a set of semi-analytical/empirical equations yielding straightforward and reliable estimates of gain and power. In a similar way, the guiding elements of a fusion reactor, to reach the prescribed fusion gain Q and power, are defined in terms of scaling relations involving pivotal quantities like radius and magnetic field. General formulae characterizing a physical system may be the consequence of an unknown symmetry. The onset of specific instabilities represent the breaking of a symmetry characterizing given equilibrium conditions. In this article, we comment on the analogy between two different physical devices, and even though we do not specify any underlying symmetry, we aim to stimulate further research in this direction. Full article