Fusion Reactor Design Research Articles

Fusion Materials Development at Forschungszentrum Jülich

Material issues pose a significant challenge for the design of future fusion reactors. From a historic point of view, the material mix used for the first wall of a fusion reactor has continually evolved, from original steel vessels to carbon and other low‐Z materials such as beryllium to tungsten as the primary candidate for a reactor's first wall armor and divertor material. For materials considered for fusion applications, a highly integrated approach is necessary. Resilience against neutron damage, good power exhaust, and oxidation resistance during accidental air ingress are design relevant issues while deciding on new materials or improving upon baseline materials. Neutron‐induced effects, e.g., transmutation adding to embrittlement, retention, and changes to thermomechanical properties, are crucial to material performance. In this contribution, the recent progress (2013–2019) in fusion materials development for current and future fusion devices, at Forschungszentrum Jülich GmbH, with activities focussing on advanced materials and their characterization is given. It is a continuation and extension of the work given by Coenen et al.

On energy confinement following the onset of ‘stiff’ transport

The dependence of confinement on input power for a tokamak plasma with regions having a stiff temperature profile is explored. The resilience of the confinement of the core energy to increasing power loss by core radiation from impurities in such situations, as it is anticipated will be required in a demonstration fusion reactor (DEMO) design, is examined.

Integrated modeling of ASDEX Upgrade plasmas combining core, pedestal and scrape-off layer physics

The design of future fusion reactors and their operational scenarios requires an accurate prediction of the plasma confinement. We have developed a new model that integrates different elements describing the main physics phenomena which determine plasma confinement. In particular, we are coupling a new pedestal transport model, based on empirical observations, to the ASTRA transport code, which, together with the TGLF turbulent transport model and the NCLASS neoclassical transport model, allows us to describe transport from the magnetic axis to the separatrix. We also coupled a simple scrape-off layer model to ASTRA, which provides the boundary conditions at the separatrix, which are a function of the main engineering parameters. By this way no experimental data of the kinetic profiles is needed, and the only inputs of the model are the magnetic field, the plasma current, the heating power, the fueling rate, the seeding rate, the plasma boundary, and the effective charge. In the modeling work-flow, first a scan in pedestal pressure is performed, by changing the pedestal width. Then the pedestal top pressure is determined using the MISHKA MHD stability code. This modeling framework is tested by simulating ASDEX Upgrade discharges. We show comparisons with experimental fueling, power, and current scans. The changes of the pedestal structure and the gradients in the plasma core, due to the different combinations of fueling, heating power, and plasma current, are well captured by the model. We also show that the predicted pedestal and global stored energies are in good agreement with the experimental measurements, and more accurate with respect to the prediction of the IPB98(y,2) scaling law. The long term goal is to obtain a robust and complete model which can be used to identify important hidden dependencies affecting global plasma confinement, which are difficult to capture by statistical regressions on global parameters.

ARC reactor materials: Activation analysis and optimization

Abstract Nowadays, Fusion Energy is one of the most important sources under study. During the last years, different designs of fusion reactors were considered. At the MIT, an innovative design was created: ARC, the Affordable Robust Compact reactor. It takes advantage of the innovative aspects of recent progress in fusion technology, such as high temperature superconductors, that permit to decrease the dimensions of the machine, reaching at the same time high magnetic fields. Our main goal is the low-activation analysis of possible structural materials for the vacuum vessel, which is designed as a single-piece placed between the first-wall and the tank that contains the breeding blanket. Due to its position, the vacuum vessel is subject to high neutron flux, which can activate it and cause the reduction of the component lifetime and decommissioning problems. The activation analysis was done also for the liquid breeder FLiBe, compared with Lithium-Lead. Codes used for the low-activation analysis were MCNP and FISPACT-II. The first one is based on a neutronics model and for each component a certain neutron flux is evaluated. For FISPACT-II, the main input is the composition of the analyzed material, the neutron flux and the irradiation time. Results from FISPACT-II are the time behavior of specific activity, contact dose rate. To assess suitable structural materials for the vacuum vessel, low-activation properties were considered. Vanadium alloys turn out to be one of the best alternatives to the present material, Inconel-718. Finally, isotopic tailoring and elemental substitution methods were applied. Here, the composition of each alloy is analyzed and critical isotopes or elements are eliminated or reduced. After the modifications, new simulations are done, and those leading to significant improvements in the final results are highlighted.

Impact of the flowing activated water on maintenance and electronic functioning for the pre-conceptual design of the European DEMO fusion reactor

This paper is devoted to provide generic guidelines to facilitate the characterization of an important contribution to the radiation environment during operation and after the shutdown. This is fundamental to plan safe operation and maintenance of the European DEMO Fusion Power Plant. The study is focused on the estimation of activation responses to electronics and workers, in representative rooms of the DEMO building housing those pipes carrying activated water coming from the water cooled breeding blankets (WCLL BB). The prediction of some radiation-related quantities is needed in order to evaluate feasibility for functioning of rad-sensible components (e.g. electronic and electrical equipment) as well as for manual maintenance planning.The feasibility for placing rad-sensible equipment near the activated water pipes is determined by calculating the accumulated dose in silicon (due to both neutrons and photons), and neutron fluence rate. Besides, the viability for manual maintenance activities is evaluated by calculating residual dose rate for workers. These quantities are assessed in a representative room containing the pipes as well as in an adjacent room. In view of future developments on DEMO designs a parametric study has been performed and a compendium of different configurations is considered: several numbers of pipes, pipes dimensions and thickness of the wall separating both rooms are examined, as well as different delay times. Additionally, a study dedicated to the shielding improvement of both rooms is presented considering such parametric variations.

The use of tungsten yarns in the production for Wf/W

Material issues pose a significant challenge for the design of future fusion reactors. Recently progress has been made towards fully dense multi short-fibre powder metallurgical production of tungsten-fibre reinforced tungsten (Wf/W) as well as optimizing the process understanding for the routes using chemical vapour deposition (CVD). For CVD-Wf/W weaves and textile preforms are being used to facilitate large scale production. Classically 150 μm tungsten fibres supplied by OSRAM GmbH have been used. In order to facilitate the better use of textile processes less stiff 16 μm filaments are being evaluated. The strength of the 16 μm filament is at 4500 MPa and thus significantly higher than the strength of the 150 μm fibre (∼2500 MPa) (in the as-fabricated state). Better weavability allows a more flexible use of fibre preforms Two main yarn production routes have been investigated: covered yarns where a set of tungsten filaments is held together by a PVA (Polyvinyl alcohol) cover and braided yarns. In order to allow a comparison to the previously used single fibres, yarns with ∼140 μm effective diameter were produced. Braided yarns with tensile strength of 2500 MPa and 6% strain at fracture and twisted yarns with tensile strength of 4500 MPa and 3% strain at fracture. For both yarns single fibre CVD samples have been produced to investigate the infiltration properties of the yarns and thus their applicability for the CVD route. A dense infiltration is observed for all yarns under investigation.

Fuel pellet injection into heavy-ion inertial fusion reactor

In inertial confinement fusion (ICF), the scientific issues include the generation and transport of driver energy, the pellet design, the uniform target implosion physics, the realistic nuclear fusion reactor design, etc. In this paper, we present a pellet injection into a power reactor in heavy ion inertial fusion. We employ a magnetic correction method to reduce the pellet alignment error in an ICF reactor chamber, including the gravity, the reactor gas drag force and the injection errors. We found that the magnetic correction device proposed in this paper is effective to construct a robust pellet injection system with a sufficiently small pellet alignment error. A fuel pellet would be covered by Pb and is cooled down by a liquid He. The cryo-pellet coated by Pb becomes a superconductor, when its temperature is below 7.2K. The magnetic field repels the superconductor due to the Meissner effect. We employ this effect to correct the fuel pellet injection velocity. The results of simulations demonstrate that the magnetic correction system is effective and robust to reduce the pellet alignment error.

Preparation and characterization of Al2O3/Y2O3 multilayer coatings as tritium permeation barrier

Tritium permeation barrier (TPB) is a promising solution to reduce fuel loss and environmental contamination caused by tritium permeation in fusion reactor. A multilayered Al2O3/Y2O3 TPB coating containing a total of 10 nanolayers was prepared by magnetron sputtering technology. The Auger electron spectroscopy, field emission scanning electron microscopy, and X-ray diffraction demonstrated the formation of the multilayer structure with crystalline Y2O3 and amorphous Al2O3 sublayers. Further, the effect of Ar/O2 flow ratio (Γ) during Al2O3 sublayer deposition on the phase structure, surface morphology, nano-hardness, corrosion resistance, He ion irradiation resistance and deuterium permeation behavior of the multilayered coating was explored systematically. The results show that the Al2O3 sublayer with lower Γ benefits the (222) preferred orientation of Y2O3 sublayer. The nano-hardness of Al2O3/Y2O3 coating gradually increases with the decrease of Γ. The coating with Γ = 1:1 performs the best corrosion resistance and He ion irradiation tolerance. The deuterium permeation test shows a lower deuterium ion current of the Al2O3/Y2O3 coating compared to their single-layer counterparts, where the deuterium gas-driven permeation process is controlled by a combination regime of recombination and diffusion. Our study could provide a potential coating material for the TPB design of fusion reactor.

Exploration of a Fast Pathway to Nuclear Fusion: Thermal Analysis and Cooling Design Considerations for the ARC Reactor

Progress in technological fields such as high-temperature superconductors, additive manufacturing, and innovative materials has led to new scenarios and to a second generation of fusion reactor designs. The new Affordable Robust Compact (ARC) fusion reactor, which compared to other designs meets its goal to achieve fusion energy in a less expensive, smaller but even more powerful, faster way, has been designed at Massachusetts Institute of Technology. In order to define ARC’s role in future electricity grids, a feasibility investigation of the load-following concept has been carried out, starting on ARC’s vacuum vessel (VV), which is the component closest to the plasma. Finite element analysis models have been designed, and thermomechanical analyses have been conducted. In this framework thermal fatigue and creep remain the main issues. This study identifies and verifies a suitable temperature range for the VV coolant. Indeed, it is found to satisfy both requirements for the lifetime of the structural material and thermodynamic efficiency optimization.

Progress Toward a Compact Fusion Reactor Using the Sheared-Flow-Stabilized Z-Pinch

The sheared-flow-stabilized (SFS) Z-pinch is a promising confinement concept for the development of a compact fusion reactor. The Z-pinch has been theoretically and experimentally shown to be stable to magnetohydrodynamic modes when sufficient radial shear of the axial flow is present. At the University of Washington, the Fusion Z-pinch Experiment (FuZE) research project examines scaling the SFS Z-pinch toward fusion conditions. The FuZE device produces long-duration, 50-cm-long pinches with measured ion and electron temperatures over 1 keV and number densities greater than cm. Plasma properties are measured with a diagnostic suite that includes magnetic field probes, heterodyne quadrature interferometry, digital holographic interferometry, ion-Doppler spectroscopy, and fast framing photography. Neutrons are produced in the FuZE device when deuterium is injected along with the normal hydrogen or helium fueling species. Neutron generation is diagnosed using plastic scintillator detectors. The neutron production is sustained for 5 to 8 μs, thousands of times longer than the static Z-pinch instability growth time. Measured neutron production is consistent with calculated theoretical values for thermonuclear yield at the observed plasma temperatures and scales with the square of the deuterium concentration. A preliminary reactor concept is designed to incorporate flowing liquid metal walls, which would serve as an electrode, a heat transfer fluid, a radiological shield, and a breeding blanket. Using a liquid metal wall could address several unresolved material and technology issues in existing fusion reactor designs.

The evolution of He nanobubbles in tungsten under fusion-relevant He ion irradiation conditions

He-induced W nanofuzz growth over the W divertor target is one of the main limiting factors affecting the current design and development of fusion reactors. In this paper, based on He reaction rate model in W, we simulate the growth and evolution of He nanobubbles during W nanofuzz formation under fusion-relevant He+ irradiation conditions. Our modeling unveils the existence of He nanobubble-enriched W surface layer (<10 nm), formed due to the He diffusion in W crystal into defect sites. At an elevated temperature, the growth of He bubbles in the W surface layer prevents He atoms diffusing into the deep layer (>10 nm). The formation of W nanofuzz at the surface is attributed to surface bursting of high-density He bubbles with their radius of ~4 nm, and an increase in the surface area of irradiated W. Our findings have been well confirmed by the experimental measurements.

The role of the source versus the collisionality in predicting a reactor density profile as observed on ASDEX Upgrade discharges

The design and optimization of a future tokamak fusion reactor, from the point of view of the plasma performance and output fusion power, requires the understanding of the underlying physics and the validation on present experiments of the theory-based tools used for the predictions. Present experimental efforts are devoted to approach reactor-relevant parameters (collisionality, , normalized heat fluxes and sources, temperatures ratio) as much as possible.In this work a series of discharges performed on ASDEX Upgrade are presented, where plasma density and auxiliary power levels are scanned to cover a parameter space that moves towards reactor relevant parameters. On this dedicated discharge dataset, a first-principle-based model is then applied for validation. The degree of agreement between code results and experimental measurements is shown and discussed. In cases where discrepancy is found, possible causes are identified.It is shown that the employed modeling tool can predict the overall trend, consistent with more fundamental theoretical considerations, although quantitative extrapolation still has to be performed with care. Key results of this work show that the density peaking is mainly sustained by turbulence, with a minor role for the fueling source, and how this experimental demonstration is important to predict future reactor density profiles. Moreover, the role of electromagnetic effects is also pointed out.

Initial Neutronics Investigation of a Liquid-Metal Plasma-Facing Fusion Nuclear Science Facility

The use of a liquid-metal (LM) plasma-facing component (LM-PFC) in fusion reactor designs has some advantages as well as some disadvantages as compared to traditional designs that use a solid plasma-facing wall. Neutronics analysis of these potential LM-PFC concepts is important in order to ensure that radiation limits are met and that system performance meets expectations.A three-dimensional (3-D) neutronics analysis parametric study considering four LM first-wall (FW) candidates, (PbLi, Li, Sn, and SnLi) was performed with a thin (2.51-cm) LM-PFC design. The 3-D neutronics study used a fusion reactor based on the Fusion Energy Systems Study (FESS) Fusion Nuclear Science Facility (FNSF) (FESS-FNSF) that served as the baseline for comparison. FESS-FNSF is a deuterium-tritium–fueled tokamak with 518 MW of fusion power. A partially homogenized 3-D computer-aided-design model of the LM-PFC FNSF design was analyzed using the DAG-MCNP5 transport code.The results show that all candidate LM designs are acceptable with 4% to 13% increases in the tritium breeding ratio compared to the baseline case. The peak displacements per atom at the FW decrease 2% to 15%. For all four LM designs examined, the magnet heating and fast neutron fluence are well below acceptable limits. Overall, the Li LM design is the best candidate from a neutronics perspective.

An analysis method and primary calorimetry verification of tokamak plasma facing components (PFCs) baking from EAST

Baking is an important wall conditioning method that is widely used in tokamak devices to obtain high quality plasma discharges. In order to create a theoretical analysis method for the baking of tokamak plasma facing components (PFCs), the COMSOL Multiphysics software is used to establish and solve a series of the heat transfer models of the experimental advanced superconducting tokamak (EAST) PFCs, as a representative calculation case. The thermohydraulic measurement system (including the temperature, pressure and flow sensors) is successfully installed on EAST to verify the accuracy of the heat transfer model, as a calorimetry verification example. The uncertainty of this analysis method as compared to the experimental results is less than 5% and the robustness of this analysis method is satisfied. These heat transfer models are used to obtain the temperatures on the surface, the internal temperature distributions of the components, and the internal temperature and velocity distributions of the cooling ducts numerically. The significance of these parameters for the fusion reactor design is discussed in detail. This analysis method is an important component of the fusion reactor thermohydraulic analysis method (FRTHAM), and it provides an accurate computing method for fusion reactor baking analysis.

Mechanical Design Concept of Superconducting Magnet System for Helical Fusion Reactor

The conceptual design of a helical fusion reactor was studied at the National Institute for Fusion Science in collaboration with other universities. Two types of the force free helical reactor (FFHR) are FFHR-d1 and FFHR-c1. FFHR-d1 is a self-ignition demonstration reactor that operates with a major radius of 15.6 m at a magnetic field intensity of 4.7 T. FFHR-c1 is a compact subignition reactor that aims to realize steady electrical self-sufficiency. Compared to FFHR-d1, FFHR-c1 has a magnetic field intensity of 7.3 T and a geometrical scale of 0.7. The location of the superconducting coils in both types of FFHR is based on that of the Large Helical Device (LHD). LHD has a major radius of 3.9 m. According to the design of LHD, the deformation must be within the required value to compensate for the accuracy of the magnetic field. According to this concept, the magnet support structure of LHD was fabricated using thick Type 316 stainless steel to impart sufficient rigidity. Thus, the stress of the magnet system of LHD is sufficiently below the permissible stress. In the case of FFHR, from the viewpoint of the reactor, a large access port is required for the maintenance of the in-vessel components. The mechanical design of the support structure is conceptualized by considering the basic thickness of the material and residual aperture space by referencing the mechanical analysis results. Details of the design concepts of LHD and FFHR-d1/FFHR-c1 as well as the results of mechanical analyses are introduced in this paper.

Application of transient CHI plasma startup to future ST and AT devices

Employment of non-inductive plasma start-up techniques would considerably simplify the design of a spherical tokamak fusion reactor. Transient coaxial helicity injection (CHI) is a promising method, expected to scale favorably to next-step reactors. However, the implications of reactor-relevant parameters on the initial breakdown phase for CHI have not yet been considered. Here, we evaluate CHI breakdown in reactor-like configurations using an extension of the Townsend avalanche theory. We find that a CHI electrode concept in which the outer vessel wall is biased to achieve breakdown, while previously successful on NSTX and HIT-II, may exhibit a severe weakness when scaled up to a reactor. On the other hand, concepts which employ localized biasing electrodes such as those used in QUEST would avoid this issue. Assuming that breakdown can be successfully attained, we then apply scaling relationships to predict plasma parameters attainable in the transient CHI discharge. Assuming the use of 1 Wb of injector flux, we find that plasma currents of 1 MA should be achievable. Furthermore, these plasmas are expected to Ohmically self-heat with more than 1 MW of power as they decay, facilitating efficient hand-off to steady-state heating sources. These optimistic scalings are supported by Tokamak Simulation Code simulations.

Effects of nitrogen-seeded deuterium plasma on tungsten surfaces

The melting and evaporation of target materials such as tungsten, due to the enormous plasma heat flux, are not compatible with the realization of ITER (International Thermonuclear Experimental Reactor) and nuclear fusion reactor designs. Nitrogen gas seeding has recently been considered as an effective radiator for edge and divertor boundary plasmas to protect the divertor plate from overheating. Although compatibility of nitrogen seeding with tokamak discharge operation was obtained in current tokamaks, the effects of nitrogen seeding on plasma-facing components, especially the divertor target material (e.g. tungsten), have not been fully tested in a fusion-relevant steady-state linear plasma system with a high plasma density and plasma heat flux, in which deuterium and nitrogen mixed-gas discharge plasmas are irradiated onto a tungsten target, closely similar to the boundary region in a magnetic fusion configuration. This study addressed such a detailed irradiation, using a variety of surface analyses. Surface-temperature-sensitive tungsten nitride formation was determined, thus elucidating the interesting surface morphology of whisker, loop, and/or hook-like nanostructures in some surface temperature bands. The surface characteristics were determined using several kinds of surface analysis methods, including x-ray diffraction, energy dispersive x-ray analysis, nanoindentation, Raman spectroscopy, and spectrometry. The possibility of tungsten contamination into facing plasmas is discussed in terms of tungsten nitride melting. The relation to industrial applications of tungsten nitride has also been mentioned.

On the energy confinement time in spherical tokamaks: implications for the design of pilot plants and fusion reactors

Experiments on NSTX and MAST have shown the thermal energy confinement time in spherical tokmaks (STs), , to have a stronger toroidal field and weaker plasma current dependence than in conventional large aspect ratio tokamaks. These scalings were derived for single machines both of which are similarly sized, consequently the NSTX and MAST scaling laws do not include a size dependence, and so cannot be used to extrapolate the performance of future STs. Using physics-based dimensional arguments we extend the NSTX scaling to include a size scaling. We also resolve a colinearity problem specific to NSTX data by assuming core transport is gyro-Bohm like. The resulting scaling has approximately zero beta dependence, a typical collisionality dependence, and a relatively weak safety factor dependence: . With the exception of the safety factor, all exponents are consistent with recent experiments in large aspect ratio tokamaks. This apparent difference between STs and large aspect ratio tokamaks is consistent with MAST and NSTX results. We have considered the implications of the scaling for pilot plants and reactors and find it may be possible to develop more compact reactors based on the ST approach.

Effect of coil configuration parameters on the mechanical behavior of the superconducting magnet system in the helical fusion reactor FFHR

FFHR-d1A and c1 are the conceptual design of a helical fusion reactor. The positional relationship among superconducting coils, a pair of helical coils with two sets of vertical-field coils, are observed to be similar in both type of FFHR. Such a relation of coil configuration is based on the coil configuration of the Large Helical Device, which has been designed and constructed at the National Institute for Fusion Science. There is increasing demand to achieve an optimized coil configuration to anticipate improvements in plasma-confinement conditions. In this study, the structural design of FFHR based on the fundamental set of parameters of coil configuration is depicted, which satisfies the soundness of the structure. Further, the effects of the coil configuration parameters on the stress distributions are investigated. An effect of radius of curvature on a winding scheme of the helical coil is also discussed.

Preparation of High Sphericity Li&lt;sub&gt;2&lt;/sub&gt;TiO&lt;sub&gt;3&lt;/sub&gt; Tritium Breeder by Polymer Assisted Sedimentation Method

The tritium breeder materials used in solid tritium breeding blanket of fusion reactor are lithium-based ceramics, like lithium titanate (Li2TiO3), lithium orthosilicate (Li4SiO4), lithium aluminate（LiA1O2) and lithium zirconate (Li8ZrO6). Among them, Li2TiO3 has the advantages of stable chemical property, high mechanical strength and high lithium content, so it has become one of the preferred materials. Based on the design of solid blanket and tritium extraction process, breeder materials need to be made into a certain shape of pebbles. B Spherical breeder is widely used in the design of fusion reactor for the larger specific surface area, more channels between the pebbles to diffuse and release tritium, and easier loading or unloading. However, due to the high requirements of the breeder pebbles, the production control are not complete. In this paper, a new preparation method of ceramic pellets with high sphericity, polymer assisted sedimentation method, was used to prepare ceramic pebbles. The slurry consists of micro-scaled ceramic powder and organic monomers. Spherical droplets of ceramic slurry were obtained by sedimentation method in viscous liquids. The organic monomer contained in the droplets was heated to polymerize into small pellet preforms during the sedimentation process. By optimizing the sintering process, the fabricated pebbles could perform high compressive load (&gt;18 N), spherical shape (0.95~1.02), uniform particle size (0.8~1.2mm) and porosity (15~20 %). The main advantages of this method is the mild preparation conditions (50~110 °C), the simplified transfer and purification process and convenient industrialization.. The obtained product has high spherical shape, uniform particle size, high compactness and mechanical strength.