Generation Of Fusion Reactors Research Articles

Frontier system-on-chip (SoC) technology for microwave diagnostics (invited).

The next generation of fusion reactors, exemplified by projects such as the Demonstration Power Plant following the International Thermonuclear Experimental Reactor, faces the monumental challenge of proving the viability of generating electricity through thermonuclear fusion. This pursuit introduces heightened complexities in diagnostic methodologies, particularly in microwave-based diagnostics. The increased neutron fluence necessitates significant reductions in vessel penetrations and the elimination of internal diagnostics, posing substantial challenges. SoC technology offers a promising solution by enabling the miniaturization, modularization, integration, and enhancing the reliability of microwave systems. After seven years of research, our team successfully pioneered the V- and W-band system-on-chip approach, leading to the development of active transmitters and passive receiver modules applied in practical settings, notably within the DIII-D tokamak project. Arrays of these modules have supported microwave imaging diagnostics. New physics measurement results from the Electron Cyclotron Emission Imaging system on DIII-D provide compelling evidence of improved diagnostics following the adoption of SoC technology. Furthermore, we achieved a breakthrough in developing an F-band SoC, advancing higher frequency capabilities for fusion devices. These achievements represent a significant leap forward in fusion diagnostic technology, marking substantial progress toward establishing reliable and efficient plasma diagnostics for future fusion reactors.

Theoretical investigation on the spectra of M1 transition of W53+ ion

The properties and spectrum of the highly charged tungsten ions draw more and more attention, due to its potential application in the next generation fusion reactor. The transition wavelength and rate of the magnetic-dipole transition between the levels of the ground configuration of W53+ ion in the range of 10−20 nm were calculated by Dirac–Fock–Slater (DFS) method with a local central potential and Multi-Configuration Dirac-Hartree-Fock (MCDHF) method, respectively. The results are consistent with the previous EBIT experimental data and other theoretical data. A collisional-radiative model was constructed, and the spectrum of magnetic-dipole transition of W53+ ion observed on EBIT (Y. Ralchenko et al., Phys. Rev. A. 83, 032517, 2011) was reproduced theoretically. Another two new transitions near the wavelength of 16.683 nm and 18.018 nm are also assigned to the M1 transition of W53+ ion.

Construction Strategy for the Future K-DEMO Superconductor Test Facility

The 16 T superconducting magnet for the next generation fusion reactor, called K-DEMO would be a great engineering challenge with current technology. The project on the test facility, named SUCCEX (Super Conducting Conductor Experiment) is prerequisite for the development of K-DEMO superconducting magnet. It would generate an external magnetic field of 16 T for developing and testing K-DEMO superconductors. Also, this project will give a guideline to fabricate future superconducting conductors for the next generation nuclear fusion reactor construction. It will be an essential facility for the production of superconducting magnets for K-DEMO. In 2021, the Korean government has approved this project and the project is in its early stages of construction. The construction strategy and the current progress are discussed in this paper.

Analysis of locally-aligned and non-aligned discretisation schemes for reactor-scale tokamak edge turbulence simulations

Edge turbulence codes will play a key role in the interpretation of ITER data and for reliable predictions of EU-DEMO. At present, such codes are not yet capable of routine simulations at reactor scale, and instead focus on smaller experiments like TCV or ASDEX Upgrade. Numerical methods have to be identified that scale to reactor size, and are able to cope with the highly anisotropic turbulent structures and simultaneously with the complex magnetic geometry in the edge with X-point(s), the vessel wall and divertor targets present. Particularly for such conditions two main approaches, non-aligned discretisation schemes and a locally-aligned discretisation scheme (commonly referred to as Flux-Coordinate Independent approach (FCI)), have emerged. We analyse both schemes concerning their applicability and scalability to next generation fusion reactors. We find that the ratio of the computational cost of a non-aligned scheme compared to aligned scheme scales as ∝(R0qρs)2 for shear Alfvén dynamics, and as (R0qρs)3 for electron heat conduction, where R0 is the major radius of a tokamak, q an estimate for the safety factor at the edge and ρs is the drift scale representing the typical size of turbulent structures to be resolved. Locally-aligned schemes can therefore be considered strongly favourable for reactor scale tokamaks concerning computational performance. On the other hand, locally-aligned schemes suffer from a more complex treatment of boundary conditions, for which an immersed boundary approach (IBA) was recently proposed. We demonstrate numerically the validity of this method in combination with the FCI approach. Finally, we present a first attempt of an ITER edge turbulence simulation with the FCI code GRILLIX at realistic parameters and in realistic geometry.

Existence of weakly quasisymmetric magnetic fields without rotational transform in asymmetric toroidal domains

A quasisymmetry is a special symmetry that enhances the ability of a magnetic field to trap charged particles. Quasisymmetric magnetic fields may allow the realization of next generation fusion reactors (stellarators) with superior performance when compared with tokamak designs. Nevertheless, the existence of such magnetic configurations lacks mathematical proof due to the complexity of the governing equations. Here, we prove the existence of weakly quasisymmetric magnetic fields by constructing explicit examples. This result is achieved by a tailored parametrization of both magnetic field and hosting toroidal domain, which are optimized to fulfill quasisymmetry. The obtained solutions hold in a toroidal volume, are smooth, possess nested flux surfaces, are not invariant under continuous Euclidean isometries, have a non-vanishing current, exhibit a vanishing rotational transform, and fit within the framework of anisotropic magnetohydrodynamics. Due to the vanishing rotational transform, these solutions are however not suitable for particle confinement.

Thermodynamics up to the melting point in a TaVCrW high entropy alloy: Systematic ab initio study aided by machine learning potentials

Multi-principal-component alloys have attracted great interest as a novel paradigm in alloy design, with often unique properties and a vast compositional space auspicious for materials discovery. High entropy alloys (HEAs) belong to this class and are being investigated for prospective nuclear applications with reported superior mechanical properties including high-temperature strength and stability compared to conventional alloys. Computational materials design has the potential to play a key role in screening such alloys, yet for high-temperature properties, challenges remain in finding an appropriate balance between accuracy and computational cost. Here we develop an approach based on density-functional theory (DFT) and thermodynamic integration aided by machine learning based interatomic potential models to address this challenge. We systematically evaluate and compare the efficiency of computing the full free energy surface and thermodynamic properties up to the melting point at different stages of the thermodynamic integration scheme. Our new approach provides a $\ifmmode\times\else\texttimes\fi{}4$ speed-up with respect to comparable free energy approaches at the level of DFT, with errors on high-temperature free energy predictions less than 1 meV/atom. Calculations are performed on an equiatomic HEA, TaVCrW---a low-activation composition and therefore of potential interest for next generation fission and fusion reactors.

The effect of neon ion irradiation on tungsten for next generation fusion reactor

Tungsten exhibits robust thermo-physical and mechanical properties in ion material interaction. It is considered as an important material of the plasma facing component for next generation fusion reactor. This material could be one of the candidates to use in the divertor region of the tokamak. With this motivation a series of experiments have been performed with a plasma focus device to study the effect of neon irradiation on tungsten in various experimental conditions like with angular variation and with number of plasma focus shots. Exposed and unirrdiated tungsten samples were analyzed subjecting to surface morphological and structural studies namely using optical microscope, scanning electron microscope, atomic force microscope, grazing incidence X-ray diffraction, Vickers hardness measurement. OM photographs are analyzed quantitatively in terms of holes and pores with angular variation and number of PF shots. Surface analyses verifies the formation of micro-cracks, bubbles, blisters, holes. X-ray diffraction pattern shows the development of compressive stress on the sample due to thermal load and formation of other phases or expanded phases.

Temperature effects on the point defects formation in [111] W by neutron induced collision cascade

Tungsten is used as plasma-facing wall in ITER where is subjected to extreme operating conditions. In this work, we study the formation damage in [111] crystalline W by neutron bombardment in the temperature range of 300–900 K which is important for designing next generation of fusion reactors. Molecular Dynamics (MD) simulations are performed at a primary knock-on atoms (PKA) energy of 1 keV within the Gaussian Approximation Potential (GAP) framework. The analysis of the induced damage is done by the Fingerprinting and Visualization Analyzer of Defects (FaVAD) which is based on a rotation invariant mapping of the local atomic neighborhoods allowing to extract, to identify and to visualize the atomic defect formation dynamically during and after the collision cascades. The evaluation allows to classify several defect types and to quantify the defect evolution as function of time and initial sample temperature. We found that the production of Frenkel pair increases as a function of the temperature due to thermal activated mechanisms.

Design, construction, and testing of no-insulation small subscale solenoids for compact tokamaks

Fusion energy systems studies (FESS) for next-step devices based on the most promising magnetic configurations indicate that high magnetic fields and high current density for magnet coil systems may reduce device size and lower the cost. High current density and radiation resistant fusion magnets are particularly beneficial for low cost, low aspect ratio compact reactor designs such as Fusion Nuclear Science Facility (FNSF), Fusion Pilot Plant (FPP) of low-aspect ratio spherical tokamak (ST) or compact stellarators. Unlike typical high field magnets for nuclear magnetic resonance (NMR), magnetic resonance imaging (MRI), or high energy physics research applications, neutron irradiation damage to organic insulations in the coil winding pack is a critical issue for next generation fusion reactors where orders of magnitude higher neutron fluence than those in present experimental reactors such as ITER are expected. Moreover, a high coil winding pack current density is needed for high field magnets in low cost, compact radial build next step fusion reactors. The slow current charging time, however, is an issue in fully non-insulated coils. We present the design, construction and testing of subscale Nb3Sn solenoids, up to half the diameter of the central solenoid (CS) for National Spherical Torus Experiment (NSTX), with and without inter-layer insulation for enhancing radiation resistance while improving the winding pack current density and coil performance. The major radius for the NSTX/NSTX-Upgrade (reference for compact ST) is 0.854/0.934 meter, and 4.8 meters for FNSF. The magnetic field for NSTX/NSTX-U at the plasma center is 0.6/1 Tesla and 9 Tesla for FNSF. The inner diameter of the central solenoid coil is 0.2/0.4 meter for NSTX/NSTX-U and 1.2 meters for FNSF. The current charging and discharging behavior of the prototype solenoids was investigated to quantify performance advantages of a simplified coil fabrication without error-prone vacuum pressure impregnation (VPI) for the removal of epoxy organic insulation. Scalability of the no insulation concept for fusion can be addressed via engineered coating for novel insulation on contact resistance or intra-layer no insulations. Radiation resistance and coil winding efficiency were significantly improved in the no insulation coils for next step compact fusion reactors.

The helium bubble: Prospects for 3He-fuelled nuclear fusion

The helium bubble: Prospects for 3He-fuelled nuclear fusion

Performance and Microstructure of Bi-2212 Wire under Different OverPressures Heat-treatment

Bi2Sr2CaCu2Ox is a potential material for the superconducting magnets of the next generation of Fusion-reactors. A R & D activity based on Bi-2212 wire is running at Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) for the feasibility demonstration of Cable In Conduit Conductor (CICC). The Bi-2212 round wires were heat-treated by the tube furnace at a pressure of 30 atm and 50 atm. Wire diameter, critical current performance, and n value were studied. The surface morphology and grain size distribution of Bi-2212 superconductor after heat treatment analyzed by Electron Backscattering Diffraction (EBSD). The performance difference of Bi-2212 at 30atm and 50atm is small, and the grain size and distribution are similar. The performance difference of Bi-2212 samples with 30atm and 50atm over-pressure heat-treatment is small. 30atm requires less equipment and can obtain a longer constant temperature zone, which is suitable for exploring the heat treatment process of a long conductor.

A density functional theory based thermodynamic model of hydrogen coverage on the W(110) surface

Tungsten will be used as a plasma facing material in the next generation of fusion reactors. To aid in understanding atomic scale H–W interactions, we investigated hydrogen coverage on the tungsten (110) surface via periodic density functional theory, providing the most stable configurations that hydrogen forms on the surface at coverage ratios of interval 0.25, step-wise, up to a full mono-layer of hydrogen. We then calculate the Gibbs free energy for the stable configurations in the presence of hydrogen gas at specified temperature and pressure. It follows that the configuration, and corresponding coverage ratio, which yields the lowest Gibbs free energy is used to estimate the macroscopic surface state. Our findings based on the model compare well to low energy electron diffraction measurements, primarily the presence of well-ordered phases with coverage ratios 0.5, 0.75, and 1.0, respectively, and that no ordered phases are expected as temperature increases and surface depletion occurs.

Molecular dynamics study of the hydrogen and helium interaction in tungsten

Tungsten (W) and W alloys are regarded as the most promising candidates for plasma facing materials (PFMs), which will be widely used in the next generation of fusion reactors. However, blistering in W-PFM induced by extremely high fluxes of low-energy hydrogen (H) and helium (He) ions irradiation will seriously influence the plasma stability and limits the lifetime of PFM. We systematically investigate the interaction between H and He interaction in W using molecular dynamics (MD) calculations. The vacancy clusters can capture a large amount of H/He, and the H-He-vacancy clusters have much higher binding energy to He than to H. The binding energy of H-He-vacancy clusters to vacancy/W is positive, and H-He-vacancy clusters have strong trapping of W atoms and vacancies. He will strongly affect the H behavior in W, He atoms and He cluster will trap H atoms, thereby preventing the diffusion of H in W.

Ionizing vs collisional radiation damage in materials: Separated, competing, and synergistic effects in Ti3SiC2

Comparison between intense radiation environments present in nuclear reactors and charged particle beams is necessary for evaluating next generation fission and fusion reactor materials. However, these two irradiation environments encompass different proportions of ionizing and collisional phenomena, so exploring the different energy loss pathways is needed for appropriate analysis. Using the candidate Mn+1AXn phase, Ti3SiC2, as a test case, this work separates the effects of electronic and nuclear energy loss during ion irradiation, through a combination of 4 MeV Au, 17 MeV Pt, and 14 MeV Cl ion irradiations to examine the effects independently and systematically recombine them. Nuclear energy loss (elastic collisions) is found to be primarily responsible for the formation of a face-centered-cubic phase with anti-site defects, while intense electronic energy loss (ionization) exacerbates the effect and increases lattice strain. Further, these dissipation pathways are found to be competing or synergistic depending on their ionization and collisional ratio.

Nature of creep deformation in nanocrystalline Tungsten

Because of Tungsten’s (W) use in high temperature and stress environments like next generation fusion reactors, exploring the creep deformation mechanism in this material is very essential. In this work, for the first time, nature of creep in nanocrystalline (nc) Tungsten and factors that govern the creep mechanism like grain size, temperature, and applied stress are studied through atomistic simulations. Applied stress is varied from 2.5 to 5.5 GPa on the metallic sample while the temperature is varied in the range of 1600–2200 K. Stress and temperature variance are repeated for average grain sizes of 2.38 nm, 2.86 nm, 3.57 nm, and 4.76 nm. From the simulations, it has been found that the creep mechanism in nanocrystalline Tungsten is contingent on the applied stress as the creep mechanism varies from lattice diffusion to grain boundary diffusion creep even up to dislocation-creep. Moreover, temperature and grain refinement seem to aid the creep phenomenon at nanoscale. For very large value of stress and temperature, the power-law fails to define the creep in nc-Tungsten. In order to quantify the propensity and mechanism of creep in nanocrystalline Tungsten, stress and grain size exponents are observed in addition to the time evolution of strain and mean square displacement. Finally, atomistic features of deformation are analyzed which evince the simulation results.

First AC loss test and analysis of a Bi2212 cable-in-conduit conductor for fusion application

The main goal of the Chinese fusion engineering test reactor (CFETR) is to build a fusion engineering tokamak reactor with a fusion power of 50–200 MW, and plan to test the breeding tritium during the fusion reaction. This may require a maximum magnetic field of the central solenoid and toroidal field coils up to 15 T. New magnet technologies should be developed for the next generation of fusion reactors with higher requirements. Bi2Sr2CaCu2Ox (Bi2212) is considered as a potential and promising superconductor for the magnets in the CFETR. R&D activities are ongoing at the Institute of Plasma Physics, Chinese Academy of Sciences for demonstration of the feasibility of a CICC based on Bi2212 round wire. One sub-size conductor cabled with 42 wires was designed, manufactured and tested with limited strand indentation during cabling and good transport performance. In this paper, the first test results and analysis on the AC loss of Bi2212 round wires and cabled conductor samples are presented. Furthermore, the impact of mechanical load on the AC loss of the sub-size conductor is investigated to represent the operation conditions with electromagnetic loads. The first tests provide an essential basis for the validation of Bi2212 CICC and its application in fusion magnets.

Experiment of joints resistance and critical current of Bi2212 conductor

Bi2Sr2CaCu2Ox is a potential material for the superconducting magnets of the next generation of Fusion reactors. A R&D activity based on Bi2212 wire is running at ASIPP for the feasibility demonstration of CICC. One sub-size conductor cabled with 42 wires was designed and manufactured. A test facility was designed and set-up to measure the resistance of joints and the critical current of the Bi2212 CICC in liquid helium. A superconducting transformer, which consisted of two concentric layer-wound superconducting solenoids, was used to provide the current of the conductor sample. Both of the primary and secondary coils were immersed in liquid helium during the experiments. The highest current of the secondary loop was up to around 14kA. The critical current of the conductor was 13.2kA with criterion of 1μV/cm and the joints resistances were lower than 50nΩ.

Metal Hall sensors for the new generation fusion reactors of DEMO scale

For the first time, the results of on-line testing of metal Hall sensors based on nano-thickness (50–70) nm gold films, which was conducted under irradiation by high-energy neutrons up to the high fluences of 1 · 1024 n · m−2, are presented. The testing has been carried out in the IBR-2 fast pulsed reactor in the neutron flux with the intensity of 1.5 · 1017 n · m−2 · s−1 at the Joint Institute for Nuclear Research. The energy spectrum of neutron flux was very close to that expected for the ex-vessel sensors locations in the ITER experimental reactor. The magnetic field sensitivity of the gold sensors was stable within the whole fluence range under research. Also, sensitivity values at the start and at the end of irradiation session were equal within the measurement error (<1%). The results obtained make it possible to recommend gold sensors for magnetic diagnostics in the new generation fusion reactors of DEMO scale.

Fabrication of sintered tungsten by spark plasma sintering and investigation of thermal stability

Tungsten has been considered as the most promising candidate for plasma-facing materials (PFMs) in a next generation fusion reactor. It is well known that commercialized ITER (International Thermonuclear Experimental Reactor) grade tungsten is manufactured by the mechanical processing at high temperature after sintering to ensure a high density with an improved structural stability. In this study, in order to obtain the high-density sintered tungsten with more enhanced structural stability, spark plasma sintering (SPS) method was employed, where a pulsed direct electric current was applied during heat treatment of powders with a pressure in the specimen. It is well known that by utilizing SPS, high-density sintered materials at a relatively lower temperature for a shorter time could be achieved compared to the other conventional sintering methods. In particular, in this study, reduction in H2 atmosphere and two-step sintering were introduced to remove the residual oxygen and achieve the full densification with suppressed grain growth at relatively low operating temperature. In an optimized condition, a fully densified sintered tungsten with a relative density of 99.9% and an average grain size of 4.4μm was fabricated. The thermal stability of tungsten specimens was evaluated by high heat flux (HHF) test, where the surface temperature was set up to 2300°C by nitrogen plasma. Then, the microstructural changes of the specimen surface have been examined after the HHF test. As a result, it was confirmed that the high-density sintered tungsten samples fabricated by SPS show an excellent microstructural stability for PFMs.

Manufacture and Test of Bi-2212 Cable-in-Conduit Conductor

The CFETR, “China Fusion Engineering Test Reactor,” is a new tokamak device. Its magnet system includes the toroidal field (TF), central solenoid (CS), and poloidal field coils. The main goal of this study is to build a fusion engineering tokamak reactor with fusion power of 50-200 MW and self-sufficiency by blanket. The maximum field of CS and TF will get around 15 T, which is much higher than that of other reactors. New materials could be used to develop the magnet technology for the next generation of fusion reactors. Bi 2 Sr 2 CaCu 2 O x is considered as a potential material for the superconducting magnets. An R&D activity is running at ASIPP for the feasibility demonstration of cable-in-conduit conductor based on the Bi2212 wire. One subsize conductor cabled with 42 wires was designed and manufactured. In this paper, the manufacturing procedures and first test results of Bi-2212 conductor samples are described in details, including wire specifications, cabling and compaction process, conductor heat treatment, as well as test results on ac loss and critical current.