Plasma-facing Wall Research Articles

The transient thermal ageing of Eurofer 97 by mitigated plasma disruptions

Plasma disruptions in a commercial-scale tokamak will impose high magnitude, short-duration thermal loads on its plasma-facing first wall. Despite a plethora of mitigation measures, these severe, off-normal events are likely to briefly expose the structural materials of the first wall to significant temperature excursions. Eurofer 97, the reference structural material for the plasma-facing first wall of the EU DEMO tokamak, is a reduced activation 9Cr steel with a normalised and tempered ferritic/martensitic microstructure. Repeated exposure to the thermal effects of disruptions over the operating lifetime of a reactor may promote a cumulative evolution of Eurofer 97′s microstructure, affecting key material properties crucial to first wall performance. This novel transient thermal degradation mechanism has been explored via a laser-based transient heating experiment, supported by time-dependent finite element thermal analysis studies of DEMO’s water-cooled lithium–lead first wall during a mitigated plasma disruption. Transient-affected samples of Eurofer 97 were characterised via scanning and transmission electron microscopy techniques (SEM/TEM), including electron backscatter diffraction (EBSD), energy-dispersive X-ray spectroscopy (EDX), and selective area electron diffraction (SAED). Microhardness and magnetisation testing data are also presented. A single 700 °C thermal transient was found sufficient to coarsen and partially recrystallise Eurofer 97′s tempered martensite sub-grains at the tungsten-Eurofer 97 interface. Further transient exposure at 700 °C resulted in the significant growth of equiaxed grains, the nucleation of intergranular Cr-rich M7C3 and M23C6 carbides, and the coarsening of V-rich and intra-granular Ta-rich MX precipitates. The microstructural effects of 850 °C transients are also reported. Notably, after 1,000 transients at 850 °C the hardness of Eurofer 97 was found to have decreased by 32 %.

Effect of substrate temperature on the structural properties of Tungsten Carbide and Tungsten-rich Tungsten Carbide films

ABSTRACT The unique mechanical properties of Tungsten Carbide (WC), such as fracture toughness, exceptional hardness, high melting temperature, outstanding thermal stability, anti-oxidation qualities fatigue and creep resistance, have made it a suitable industrial material for different applications. It is also considered to be a promising material for coating on plasma-facing walls in a Tokamak reactor. The aim of the present work is to study the effect of varying substrate temperature, concentration of tungsten (W) and annealing in the structural properties of WC films. The WC films are deposited on a silicon substrate at different substrate temperatures viz. RT (300 K), 400, 500, 600 and 700 K using DC magnetron sputtering. W-rich WC films were deposited keeping the power of WC constant (100 Watt) and varying W powers (10 Watt, 20 Watt, 30 Watt, 40 Watt) at room temperature. Later, these W-rich WC films were annealed at 800°C in a reducing atmosphere. Glancing incidence X-ray diffraction (GIXRD) and Raman spectroscopy of the films are performed for structural analysis. GIXRD of the WC films reveals that as-deposited RT films are amorphous. However, films which are either deposited at elevated substrate temperature or are annealed are crystalline. This is confirmed by Raman’s measurements. It also indicates the presence of disorder and nanocrystalline phases due to temperature variations.

Compact fusion blanket using plasma facing liquid Li-LiH walls and Pb pebbles

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 1019 n/cm2, and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH2), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements.

Key technologies of manufacturing and testing for the ITER blanket shield block

The SB is one of the main in-vessel components manufactured with SS316L(N)-IG to play a major role in neutron shielding, to support the FW panels and to supply the FW panel with cooling water. Cooling channels are located inside of the SB to remove the neutron heat deposition while keeping the structure temperature to acceptable levels. The water coolant is routed first through the plasma-facing first wall panel which is attached to the SB and then through the SB. A number of deep slits are introduced into the SB to disrupt eddy current loops and reduce the Electro-Magnetic (EM) loads on the support system and VV. Water headers are machined on the side of the module with 7–10 mm welded cover plates. A Procurement Arrangements (PAs) for the manufacturing and testing of the SBs have been signed with Korea (50 %) and China (50 %).As part of the PA in Korean Domestic Agency (KODA), Process Qualification (PQ) to ensure the manufacturing and testing ability of the SB was completed successfully through manufacture of a Full Scale Prototype (FSP) as an essential step before starting series production. This paper reports key technologies, such as machining, welding, NDE and hot helium leak testing, of manufacturing and testing for the ITER SB series production in Korea toward successful completion of the missions.

The integrated engineering design concept of the upper limiter within the EU-DEMO LIMITER system

The EU-DEMO first wall protection relies on a system of limiters. Although they are primarily designed for facing the energy released by a limited plasma during transients, their design should safely withstand a combination of loads relevant for in-vessel components (IVCs) during steady-state operation. They are not meant to breed tritium, nor to provide plasma stability. However, sitting in place of blanket portions, they should ensure an adequate shielding function to vacuum vessel and magnets while withstanding both their dead weight and the electro-mechanical loads arising from the interaction between current induced in the conductive structure and magnetic field. During plasma disruptions they will be subjected to halo currents flowing from/to the plasma and the grounded structures, whose effects must be added to the eddy current ones. Disruption-induced electro-mechanical loads are hence IVC design-driving, despite the uncertainties in both eddy and halo currents’ magnitude and distribution, which depend on IVC design, electrical connectivity, plasma temperature and halo width.The integrated design of the limiter is made of two actively water-cooled sub-components: the Plasma-Facing Wall (PFW) directly exposed to the plasma, and the Shielding Block (SB) devoted to hold the PFW while providing neutronic shielding. The PFW design is driven by disruptive heat loads. Disruption-induced electro-magnetic loads are instead SB design drivers, meaning that the design details (i.e. geometry, electrical connections, attachments) affect the loads acting on it, which, in turn, are affected by the mechanical response of the structure.The present paper describes the design workflow and assessment of the Upper Limiter (UL), resulting from a close and iterative synergy among different fields. Built on static-structural and energy balance hand calculations based on, respectively, preliminary electro-magnetic and neutronic loads, the UL integrated design performance has then been verified against electro-magnetic, neutronic, thermal-hydraulic and structural assessment under the above-mentioned load combination. The outcome will be taken as reference for future limiter engineering designs.

Practical Model for the Calculation of Lateral Electromagnetic Loads in Tokamaks at Asymmetric Vertical Displacement Events (AVDEs)

This paper describes a new practical numerical model for the calculation of lateral electromagnetic (EM) loads in tokamaks during asymmetric vertical displacement events (AVDEs). The model combines key features of two recently reported trial models while avoiding their drawbacks. Their common basic feature is the superposition of two patterns of halo current: one perfectly symmetric and another perfectly anti-symmetric. This model combines the following features that have not been combined before (a) a helically distorted halo layer wrapping around core plasma, and (b) halo-to-wall interception belts slipping along plasma-facing walls. This combination almost doubles the lateral net forces. An AVDE creates significant lateral net moments. Being relatively modest at VDEs, the lateral moments become a dominant component of EM loads at AVDEs. The model carefully tracks the balance of net EM loads (zero total for the tokamak), as a necessary condition for the consequent numerical simulation of the tokamak’s dynamic response. This balance is needed as well for the development of tokamak monitoring algorithms and simulators. In order to decouple from the current uncertainties in the interpretation and simulation of AVDE physics, the model does not simulate AVDE evolution but uses it as an input assumption based on the existing interpretation and simulation of AVDE physics. This means the model is to be used in a manner of parametric study, at widely varied input assumptions on AVDE evolution and severity. Parametric results will fill a library of ready-for-use waveforms of asymmetric EM loads (distributed and total) at tokamak structures and coils, so that the physics community may point to specific variants for subsequent engineering analysis. This article presents the first practical contribution to this AVDE library.

The behavior of helium bubble evolution under neutron irradiation in different tungsten surfaces

Tungsten, as a kind of material for the plasma-facing wall in fusion reactor and tokamak, is subjected to irradiation by the high-energy neutron and the high fluxes of hydrogen and helium plasmas. Especially, in the presence of helium (He) bubbles, neutron irradiation can significantly exacerbate the deterioration of the material's mechanical properties. However, due to experimental limitations, the evolution of He bubbles in tungsten irradiated by neutrons and the underlying mechanisms remain unclear. Here, molecular dynamics (MD) simulation is used to investigate the behavior of He bubbles in tungsten with different crystal surfaces irradiated by neutrons. We observed the complete evolutionary process of the He bubble, i.e. standing, expanding, and bursting. During this process, high-energy neutron transfers energy to atoms within the He bubble through cascade collision, leading to a rapid increase in energy and pressure of the He bubble, and then the rapid expansion of the He bubble makes the pressure rapidly released, ultimately leading to the rupture of the helium bubble. Furthermore, we also explore the evolution of the He bubble under varying conditions (such as temperature, neutron energy, and He/vacancy ratio) within the same crystal surface. We found that at lower temperatures and neutron energy with a higher He/vacancy ratio, it tends to form pinhole-like channels to release helium atoms, resulting in the He bubble bursting. On the contrary, it tends to form a hill-like expansion to spurt helium atoms out, leading to the bubble bursting. Additionally, it was found that the direction and time of He bubble rupture varied under different crystal surfaces. He bubbles under the (112) surface burst the slowest with excellent corrosion resistance. These findings provide new insights into the behavior of irradiation resistance of fusion reactor materials under extreme conditions and offer significant guidance for the improvement and selection of optimal materials for plasma-facing walls.

Main Nuclear Responses of the DEMO Tokamak with Different In-Vessel Component Configurations

Research and development of the DEMOnstration power plant (DEMO) breeder blanket (BB) has been performed in recent years based on a predefined DEMO tritium breeding ratio (TBR) requirement, which determines a loss of wall surface due to non-breeding in-vessel components (IVCs) which consume plasma-facing wall surface and do not contribute to the breeding of tritium. The integration of different IVCs, such as plasma limiters, neutral beam injectors, electron cyclotron launchers and diagnostic systems, requires cut-outs in the BB, resulting in a loss of the breeder blanket volume, TBR and power generation, respectively. The neutronic analyses presented here have the goal of providing an assessment of the TBR losses associated with each IVC. Previously performed studies on this topic were carried out with simplified, homogenized BB geometry models. To address the effect of the detailed heterogeneous structure of the BBs on the TBR losses due to the inclusion of the IVCs in the tokamak, a series of blanket geometry models were developed for integration in the latest DEMO base model. The assessment was performed for both types of BBs currently developed within the EUROfusion project, the helium-cooled pebble bed (HCPB) and water-cooled lead–lithium (WCLL) concepts, and for the water-cooled lead and ceramic breeder (WLCB) hybrid BB concept. The neutronic simulations were performed using the MCNP6.2 Monte Carlo code with the Joint Evaluated Fission and Fusion File (JEFF) 3.3 data library. For each BB concept, a 22.5° toroidal sector of the DEMO tokamak was developed to assess the TBR and nuclear power generation in the breeder blankets. For the geometry models with the breeder blanket space filled only with blankets without considering IVCs, the results of the TBR calculations were 1.173, 1.150 and 1.140 for the HCPB, WCLL and WLCB BB concepts, respectively. The TBR impact of all IVCs and the losses of the power generation were estimated as a superposition of the individual effects.

Appearance and analysis of a reflecting coating damage

Eight reflecting gratings are installed into the plasma facing wall of ASDEX Upgrade (AUG) in order to provide a controlled second pass through the plasma centre in 140 GHz heating scenarios with reduced single pass absorption. Four of these gratings are machined out of W1.4901 steel and coated with tungsten to increase the reflectivity. During plasma operation three of them worked very well, only one showed a strong correlation between the launcher On-time and an unusual increase in plasma radiation. After completion of the 2022 experimental campaign, this tile was carefully inspected. Traces of local melting were visible and the tile was examined with a scanning electron microscope to determine the surface material composition. The image of backscattered electrons revealed that tungsten is missing locally and along some of the ridges of the complex topology of this grating. Within these areas, the steel surface started to melt, which is in accordance with the assumption, that an intact tungsten coating indeed prevents the steel from melting. The damaged tile is currently being replaced and we have implemented two measures in order to prevent such damage on the new tile. The first measure is to consequently finish all machining steps before the coating procedure. This is because a mechanical damage of the coating before the installation could not be ruled out. The second measure is to control and minimize the surface roughness after machining and before the coating procedure. It turned out that the roughness was up to 3 microns in the past, which seems to be too high for the desired quality of this particular coating. We have tested and developed an electropolishing procedure to decrease the surface roughness to the order of 1 micron and keep the grating topology as precise as possible.

The dependence of tungsten fuzz layer thickness and porosity on tungsten deposition rate and helium ion fluence

Fuzz formation on a heated tungsten surface in the presence of a helium-containing plasma and tungsten deposition source was investigated. Tungsten samples were exposed at 1123 K to pure helium plasma with ion incident energy of 76 eV, W/He ion flux ratio of ∼0.4×10−4 , and varied helium ion fluence from 0.18 to 3.4×1026 m−2. Fuzz thickness was measured by cross-sectional scanning electron microscopy to increase from 0.22 to 15 µm with increasing helium ion fluence. No indication of saturation in fuzz thickness at high fluence was observed, in contrast to fuzz produced on a tungsten surface without tungsten deposition. Additional tungsten samples were exposed at 1123 K to pure helium plasma with ion incident energy of 76 eV, helium ion fluence of ∼3.4×1026 m−2, and varied W/He ion flux ratio from 0.26 to 3.0×10−4 . Fuzz thickness increased from 7.5 to 120 µm with increasing W/He ion ratio. A final sample exposed at 1123 K to a mixed helium-deuterium plasma with ion incident energy of 76 eV, helium ion fluence of 0.18×1026 m−2, and W/He ion flux ratio of 2.2×10−4 developed nearly identical fuzz structures to that developed in a pure He plasma. As a function of deposited tungsten fluence, all results were found to trace out a single layer-growth curve given by a power law relation, indicating that fuzz thickness is independent of the W/He ion flux ratio in the range investigated and independent of any deuterium present in the plasma. As a result, for tungsten plasma facing walls in magnetic fusion devices at 1000–2000 K with 10−4 W/He ion flux ratio, fuzz with thicknesses greater than hundreds of microns may form in as little as 104 s (in the absence of ELM-induced erosion or annealing), and may more significantly affect its thermophysical properties than fuzz generated without a tungsten deposition source.

Tungsten based divertor development for Wendelstein 7-X

Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified.This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx. < 0.25 m2 and < 60 kg) using the existing water cooling infrastructure providing 5 l/s and roughly maximum 15 bar pressure drop per module.The main technology under qualification is based on a CuCrZr heat sink made either by additive manufacturing using laser powder bed fusion (LPBF) or by uniaxial diffusion welding of pre-machined forged CuCrZr plates. After heat treatment, the plasma-facing side of the heat sink is covered by W or if feasible by the more ductile WNiFe, preferably by coating or alternatively by hot isostatic pressing W based tiles with a soft OFE-Cu interlayer. Last step is a final machining of the plasma-exposed surface and the interfaces to the water supply lines and supports to correct manufacturing deformations.

The investigation of the neutronic responses of components in the K-DEMO

The neutronic response of consisting materials of the Korean fusion DEMOnstration reactor (K-DEMO) in its fusion neutron environment is a crucial consideration factor from the conceptual design stage of the K-DEMO. Especially in the design of in-vessel components (IVCs) of the K-DEMO that are placed in the most extreme neutron irradiation field, neutronic damage of constituent components is a major limiting factor that determines the lifetime of IVCs. From the analysis of the neutronic response of IVCs of the K-DEMO, the neutron wall loading (NWL) related to the tritium breeding ratio and nuclear heating of IVCs can be quantified to assess the self-sufficient supply of tritium, and thermal energy transferred from fusion neutrons, respectively. The calculated NWL shows that the harmonizing design of the cooling configuration of each blanket segment with the corresponding NWL is critical thermos-hydraulic design issue for the efficient utilization of thermal energy in the blanket. Another finding is that the double null magnetic field configuration and related blanket configuration with a water-cooled pebble bed of the K-DEMO make a self-sufficient tritium supply challenging. The implicated lifetime of the first plasma-facing tungsten wall of the K-DEMO is around 2 full power years (FPY) in the severely neutron-irradiated region. For the reduced activation ferritic/martensitic steel layers in the blanket, the lifetime of it is estimated around 4 FPY in the inboard region. Based on the response analysis results of this study, optimization of the design of the K-DEMO will continue iteratively in the future.

Polarized fusion and potential in situ tests of fuel polarization survival in a tokamak plasma

The use of spin-polarized fusion fuels would provide a significant boost towards the ignition of a burning plasma. The cross section for D + T → α + n, would be increased by 1.5 if the fuels were injected with parallel polarization. Furthermore, our simulations demonstrate additional non-linear power gains in large-scale machines such as ITER, due to increased alpha heating. Such benefits require the survival of spin polarizations for periods comparable to the particle confinement time. During the 1980s, calculations predicted that polarizations could survive a plasma environment, although concerns persisted regarding the cumulative impacts of wall recycling. In that era, technical challenges prevented direct tests and left the large scale fueling of a power reactor beyond reach. Over the last decades, this situation has changed dramatically. Detailed simulations of ITER have predicted negligible wall recycling in a high-power reactor, and recent advances in laser-driven sources project the capability of producing large quantities of ∼100% polarized D and T. The remaining crucial step is an in-situ demonstration of polarization survival in a plasma. For this, we outline a measurement strategy using the isospin-mirror reaction, D + 3He → α + p. Polarized 3He avoids the complexities of handling tritium, while encompassing the same spin-physics. We evaluate two methods of delivering deuterium, using dynamically polarized Lithium-Deuteride (with vector polarization PV D of 70%) or frozen-spin Hydrogen-Deuteride (with PV D of 40%), together with a method of injecting optically-pumped 3He (with 65% polarization). Pellets of these materials all have long polarization decay times (∼6 min for LiD at 2 K, ∼2 months for HD at 2 K, and ∼3 d for 3He at 77 K), all far greater than a plasma shot in a research tokamak such as DIII-D (∼20 s). Both species can be propelled from a single cryogenic injection gun. We review plasma requirements and strategies for detecting polarization survival. Polarization alters both fusion yields and the angular distribution of fusion products, and each of these provides a potential signal. In this paper we simulate a selection of shots with similar characteristics in a future high-Tion H plasma, and find ratios of yields from shots with fuel spins parallel and antiparallel reaching 1.3 (HD + 3He) to 1.6 (LiD + 3He) over a wide range of poloidal angles. (A companion paper finds sensitivity to fusion product angular distributions as reflected in the pitch angles of protons and alphas reaching the plasma facing wall.)

Virtual prototyping of liquid metal blanket performance in fusion pilot plant

Liquid metal blanket is a dominant design option for the next step fusion devices responsible for harvesting energy from fusion reaction, and simultaneously producing fuel for the same reaction through tritium breeding. Liquid metal blankets introduce additional complexity to the design due to fluid motion, fluid structure interaction, and magnetohydrodynamic (MHD) effects arising from the motion of the conducting fluid through the magnetic field. They are also directly affected by the plasma heat flux and neutronic fluence. PPPL is currently developing a virtual prototyping system for numerical analysis of the liquid metal blankets for future fusion devices. The system has a customized 3D computational fluid dynamics (CFD) code in its core, allowing MHD flow and conjugate heat transfer analysis in blankets fluids and solids. The code was successfully used before for dual coolant blanket analysis [A. Khodak et al., Fusion Eng. and Des. 137 (2018)]. Recently the same code was modified to allow verified simulation of MHD flows at high Hartmann numbers of several thousand typical for blanket applications. CFD code receives volumetric heat source distribution from the neutronic analysis based on MCNP code. In addition, direct tritium breeding simulation will be performed allowing optimization of the blanket performance. 2D axisymmetric version of neutronics code will be used for rapid optimization, with 3D version employed for detailed analysis. The surface heat distribution on the plasma facing wall will be defined by the software HEAT allowing 3D modeling of the heat flux based on the magnetic field distribution including gyro-orbit effects. Results of thermal analysis are imported into structural analysis code also included in the system. Direct import of CAD geometry will be used for analyzing all components and as a result design option can be efficiently optimized.

Wall conditioning effects and boron migration during boron powder injection in ASDEX Upgrade

The efficacy of boron powder injection as a wall conditioning tool method in terms of its ability to create a sufficiently uniform boron layer on plasma-facing surfaces has been studied in ASDEX Upgrade. Boron powder was injected in two series of dedicated plasma discharges at varied injection rates and total amount injected. The resulting boron deposition was determined quantitatively by exposure of witness samples followed by ex-situ surface analysis of the retrieved samples. In both experiments, isotopically enriched boron was used to distinguish the deposition of the newly injected material from the residual boron fraction in the machine originating from previous glow discharge boronisations. It could be confirmed that the injected boron is migrating and re-deposited across plasma-facing wall surfaces already within one discharge. At erosion dominated divertor areas, the boron influx from the main chamber results in formation of a mixed tungsten-boron surface layer with a boron area density of O(1nm) whereas at deposition dominated areas closed boron layers grow with ongoing boron injection to a thickness of up to O(1μm). Extrapolating the radial boron deposition profile on the samples exposed in the main chamber to the limiter front yields a similar boron coverage of O(1μm), which is about ten times higher than typical values for glow discharge boronisation. Together with the observed reduction of oxygen level and improved wall pumping, the surface analysis results demonstrate that boron powder injection provides a suitable means to refresh the wall conditioning effect of a preceding glow discharge boronisation.

Microstructural modelling and characterisation of laser-keyhole welded Eurofer 97

The novel reduced activation ferritic/martensitic steel Eurofer 97 is employed by many concept designs for the plasma-facing first wall of the EU DEMO fusion reactor. These designs feature precision joints between Eurofer 97 coolant piping, for which an advanced laser-keyhole welding technique is proposed. In this work the microstructure of these novel laser-keyhole Eurofer 97 welds is modelled by combining finite element thermal analysis with precipitate kinetics modelling. Microanalysis of a representative specimen via scanning electron and high-speed atomic force microscopy techniques is also conducted, complimented by electron backscatter diffraction, energy-dispersive X-ray spectroscopy, and nanoindentation hardness testing. Models of the weld microstructure agree well with the results of microanalysis although the precipitate diameters predicted are slightly underestimated. Several large void defects were discovered within the weld fusion zone, the cause of which is suspected to arise from the evaporation of cerium-rich oxide inclusions present in the as-cast Eurofer 97 during welding.

An optimized power conversion system for a stellarator-based nuclear fusion power plant

Fusion is a candidate base-load and load-following energy source for the future carbon-free energy mix, and stellarators are promising magnetic fusion reactors. Unlike all other fusion devices, which are pulsed, the stellarator is inherently steady-state. Thus it requires a different downstream power conversion system compared to tokamaks. In this paper, we conceive and optimize such a system for a stellarator power plant equipped with plasma-facing liquid metal walls in the 700–900 °C temperature range. We rely on a supercritical CO2 Brayton–Rankine Combined Cycle, and optimize the power plant via a genetic algorithm. We also compare the effect of the characteristic parameters on the stellarator performance. The efficiency of the power conversion system is 51% and net electrical efficiency of the complete plant (including reactor auxiliaries) is 34%. Such figures are remarkably higher compared to the state-of-the-art designs in the field of thermo-nuclear fusion plants, considering that the most optimistic and recent estimate foresee a heat-to-power conversion efficiency of about 34% for tokamak-like reactors. Finally, we discuss the technical feasibility of the two most critical components for operations: (i) the large-scale supercritical CO2 Gas Turbine, and (ii) the compact heat exchanger for the Brayton cycle. We obtain compact machines since the maximum diameter calculated for the s-CO2 compressor is about 1.2 m with a rotational speed of 3000 rpm, and the volume required to accommodate the heat exchange surface could achieve 40 m3 with an area density of 7900 m2/m3.

Computational Assessment on Impurity Sourcing and Transport Using High-Temperature Graphite and Silicon Carbide Plasma-Facing Walls in a Tokamak Environment

Operating with hot tokamak plasma-facing components will be essential in fusion reactors to maximize the thermal efficiency of the blanket. The SOLPS-ITER edge plasma code package and the DIVIMP Monte Carlo impurity tracking code were used in tandem to simulate the effect of active wall heating on impurity sourcing and transport in a DIII-D–size tokamak. The SOLPS-ITER plasma background was generated based on a previous DIII-D discharge and includes the effect of particle drifts. DIVIMP simulations found that actively heating the lower divertor (versus the divertor shelf or the entire wall) was the most efficient way to minimize gross erosion and core impurity influx at temperatures above 1000 K. Replacing the graphite wall with a silicon carbide (SiC) wall yielded a 5 to 20× decrease in the estimated gross erosion rate of carbon, with a maximum decrease observed at a lower divertor temperature of 800 K. Gross erosion of Si from SiC was estimated to be almost 100× lower than that of C from SiC, due primarily to the low impact energy of incident D plasma on the divertor targets. The core impurity influx for SiC walls is predicted to be lower than that with graphite walls, but eroded Si ions appear to migrate preferentially (versus C) to the core due to a more peaked erosion profile closer to the strike points where the ion temperature gradient force drives particles upstream. These predictive simulations suggest that active heating of the plasma-facing wall may both lower wall erosion and improve core performance relative to the “warm” walls of current devices that are typically only heated via plasma contact. Relative reductions in gross erosion and upstream accumulation by using SiC instead of graphite as the wall material strengthen the argument for upgrades to current graphite-clad machines and continued development of SiC first-wall and blanket concepts.

Plasma facing components with capillary porous system and liquid metal coolant flow

Liquid metal can create a renewable protective surface on plasma facing components (PFC), with an additional advantage of deuterium pumping and the prospect of tritium extraction if liquid lithium (LL) is used and maintained below 450 °C, the temperature above which LL vapor pressure begins to contaminate the plasma. LM can also be utilized as an efficient coolant, driven by the Lorentz force created with the help of the magnetic field in fusion devices. Capillary porous systems can serve as a conduit of LM and simultaneously provide stabilization of the LM flow, protecting against spills into the plasma. Recently, a combination of a fast-flowing LM cooling system with a porous plasma facing wall (CPSF) was investigated [A. Khodak and R. Maingi, Nucl. Mater. Energy 26, 100935 (2021)]. The system takes an advantage of a magnetohydrodynamics velocity profile as well as attractive LM properties to promote efficient heat transfer from the plasma to the LL at low pumping energy cost, relative to the incident heat flux on the PFC. In the case of a disruption leading to excessive heat flux from the plasma to the LM PFCs, LL evaporation can stabilize the PFC surface temperature, due to high evaporation heat and apparent vapor shielding. The proposed CPSF was optimized analytically for the conditions of a fusion nuclear science facility [Kessel et al., Fusion Sci. Technol. 75, 886 (2019)]: 10 T toroidal field and 10 MW/m2 peak incident heat flux. Computational fluid dynamics analysis confirmed that a CPSF system with 2.5 mm square channels can pump enough LL so that no additional coolant is needed.

Limiters for DEMO wall protection: Initial design concepts & technology options

Off-normal macroscopic plasma instabilities which possibly occur in a fusion power plant represent the most critical operational accidents and pose a serious risk to economically viable exploitation of fusion power. During such an instability, a huge amount of plasma energy is rapidly deposited onto the plasma-facing first wall in form of extreme heat flux. Such heavy thermal shocks can cause severe damages or a fatal failure of the first wall even by a single event. This issue is a serious design concern raising the requirement of a wall protection strategy. To cope with this issue, the European DEMO project adopted the limiter concept. The limiters equipped with tungsten sacrificial armor shall be installed at those wall areas where plasma contact is expected to be probable. To implement the limiter concept, comprehensive engineering studies have been conducted including conceptual design of the plasma facing component and technology R&D for armor material and joining. In this overview paper, the background of the wall protection issue and its implication on design and materials are explained and the outcomes from the recent project period (2019-20) are reported. The focus is placed on the technology achievement related to the development of novel lattice-type tungsten armor material fabricated by an innovative additive manufacturing process. Interim results from the extreme and medium heat flux tests and the joining trials for tungsten-to-heat sink material (copper or steel) are also addressed. The design rationale and the interpretation of the test data are supported by computational simulations.