Fusion Sci Research Articles

Bayesian parameter estimation and evaluation of the K-ω shear stress transport model for plane impinging jets

Numerical simulations with semi-empirical turbulence models are commonly used to model impinging jets, often used for cooling solid surfaces. In this work, the constants in the k-ω shear stress transport model in ANSYS FLUENT are calibrated to experimental velocity and heat transfer data for a plane turbulent impinging air jet to determine if Kennedy-O’Hagan calibration (Kennedy and O’Hagan 2001 J. R. Stat. Soc. B 63 425–64) can improve predictions of near-surface velocities and surface Nusselt numbers for similar flows. Impinging jets have been proposed to cool the target plates of the divertor in future magnetic fusion energy reactors, where simulations are used to estimate divertor performance. The flat-plate divertor (Wang et al 2009 Fusion Sci. Technol. 56 1023–7) uses a plane jet of helium issuing from a B = 0.5 mm slot to cool a surface with radius of curvature of 44B at a distance 4B from the slot. Predictions from the calibrated numerical model are compared with independent experimental data at different flow conditions, as well as surface temperature data for a flat plate divertor test section. The contribution of this work is evaluation of the accuracy of a calibrated turbulence model for modest extrapolations in flow geometry and flow conditions for a plane impinging jet.

Modeling ablator defects as a source of mix in high-performance implosions at the National Ignition Facility

Recent indirect drive inertial confinement fusion implosions on the National Ignition Facility (NIF) [Spaeth et al., Fusion Sci. Technol. 69, 25 (2016)] have crossed the threshold of ignition. However, performance has been variable due to several factors. One of the leading sources of variability is the quality of the high-density carbon (HDC) shells used as ablators in these experiments. In particular, these shells can have a number of defects that have been found to correlate with the appearance of ablator mix into the hot spot and a degradation in nuclear yield. These defects include pits on the ablator surface, voids in the ablator bulk, high-Z debris from the Hohlraum wall that adheres to the capsule surface, and finally the inherent granular micro-structure of the crystalline HDC itself. This paper summarizes high-resolution modeling of each of these mix sources in two recent high-performance NIF implosion experiments. The simulated impact from a range of individual capsule defects is found to be broadly consistent with the trends seen in experiment, lending credence to the modeling results and the details of the mixing process that they reveal. Interestingly, modeling of the micro-structure inherent to HDC shows that this perturbation source results in considerable mixing of the deuterium–tritium fuel with ablator material during the implosion. The reduction in fuel compression from this mix results in an approximately factor of two reduction in neutron yield in current implosions and emphasizes the importance of mitigating this significant performance degradation.

Plasma profile reconstruction supported by kinetic modeling

Combining the analysis of multiple diagnostics and well-chosen prior information in the framework of Bayesian probability theory, the Integrated Data Analysis code (IDA Fischer et al 2010 Fusion Sci. Technol. 58 675–84) can provide density and temperature radial profiles of fusion plasmas. These IDA-fitted measurements are then used for further analysis, such as discharge simulations and other experimental data analysis. Since IDA considers measurement data, which is frequently fragmentary, with statistical and systematic uncertainties, which are often difficult to quantify, from a heterogeneous set of diagnostics, the fitted profiles and their gradients may be in contradiction to well-established expectations from transport theory. Using the modeling suite ASTRA coupled with the quasi-linear transport solver TGLF, we have created a loop in which simulated profiles and their uncertainties are fed back into IDA as an additional prior, thus providing constraints about the physically reasonable parameter space. We apply this physics-motivated prior to several different plasma scenarios and find improved heat flux match, while still matching the experimental data. This work feeds into a broader effort to make IDA more robust against measurement uncertainties or lack of measurements by combining multiple transport solvers with different levels of complexity and computing costs in a multi-fidelity approach.

Modeling transient edge plasma transport with dynamic recycling

The work presents numerical simulation studies of the role that dynamic plasma recycling on the main wall and divertor target surfaces plays in transient edge plasma transport phenomena, such as edge localized modes (ELMs). The studies are performed by coupling the edge plasma transport code UEDGE [Rognlien et al., J. Nucl. Mater. 196–198, 347 (1992)] and the wall reaction–diffusion transport code FACE [Smirnov et al., Fusion Sci. Technol. 71, 75 (2017)]. The two-dimensional, time-dependent, two-way coupling of the codes, in a realistic tokamak geometry, is accomplished using the Integrated Plasma Simulator framework [Elwasif et al., in 18th Euromicro Conference on Parallel, Distributed and Network-Based Processing (PDP 2010), Pisa, Italy (IEEE, 2010), pp. 419–427] for all modeled material plasma boundaries. The simulations show that dynamic plasma recycling has substantially different characteristics on the main wall and on the divertor plates. It is demonstrated that during an ELM cycle the outer wall can dynamically absorb and release a number of particles comparable to that expelled by the ELM from the core plasma, by far exceeding the dynamic retention capacity of the divertor surfaces. The resulting evolution of the edge and divertor plasma conditions during an ELM cycle is analyzed.

Temperature and density dependence of Kr L-shell spectrum in hot dense plasmas

Kr L-shell spectroscopy modeling results are discussed in this paper, focusing on the n = 4 to n = 2 line transitions of Be- and Li-like Kr ions. Collisional radiative atomic kinetic and Stark-broadened spectral line shape calculations show electron temperature Te and density ne sensitivity in the spectrum. The combination of the Te dependence due to the relative intensity of Be-like to Li-like line emissions in the range from 1.5 to 3 keV and the ne sensitivity from the Stark broadening effect on the line shapes in the range from 5×1023 to 2×1024/ cc results in a spectrum that can be employed to diagnose Te and ne. Two different collisional radiative atomic kinetic models i.e., Prismspect [J. J. MacFarlane, et al., Int. Fusion Sci. Appl. Conf. Proc. 457 (2003)] and ABAKO [Florido, et al., PRE, 80, 056402 (2009)] produce similar results in level populations and spectra. In x-ray spectroscopy of implosion cores, this Kr L-shell spectrum may prove useful in an intermediate Te range in which Ar is too ionized for its K-shell to be of diagnostic value and Kr is not ionized enough for its K-shell emission to be useful.

Theory based recommendations to the resistive wall mode stability studies in tokamaks

The problem of the plasma stability against the resistive wall modes is considered from two sides, theoretical and experimental. The main subject is the dispersion relation and its verification, which is commonly understood as a comparison of the predicted and observed stability thresholds. As in the conventional magnetohydrodynamics, the growth rate γ and the angular rotation frequency ω of the mode are found from the energy balance with account of some dissipation in the plasma, additional to the resistive losses in the vacuum vessel wall. The resulting relations are integral, which allows the same γ and ω with different integrands. It is shown explicitly that only two fitting parameters are needed for getting a perfect agreement of such results with measured γ and ω. This explains why all attempts with so-called kinetic relations have been good in that. This also reveals the reason for the earlier finding [A. M. Garofalo, Fusion Sci. Technol. 48, 918 (2005)] that a number of models provided the stability regardless of the type of dissipation as long as the dissipation was sufficiently large. It is shown here that such “degeneracy” is a general property. One consequence is that a similar success with any model cannot guarantee its validity, and none of them can be recommended to ITER immediately. It is also explained that the edge harmonic oscillations can be a promising candidate for testing the dissipation channels missing in the kinetic dispersion relations.

W-band tunable, multi-channel, frequency comb Doppler backscattering diagnostic in the ASDEX-Upgrade tokamak

This article presents the design, implementation, and first data of a uniquely flexible, multi-channel, frequency comb Doppler backscattering diagnostic recently made operational in the ASDEX-Upgrade tokamak [A. Gruber and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)]. It uses a double side-band signal fed into a ×6 frequency multiplier to produce a multiple-frequency output spectrum. Seven of these frequencies are simultaneously measured in the receiver via a two-step frequency down-conversion and traditional I/Q demodulation. The frequency comb spectrum is fully tunable to sit anywhere in the W-band. The inter-frequency separation is also uniquely tunable remotely between 0.1 and 6 GHz without any hardware changes. The diagnostic can be operated in both O and X-mode polarizations and at both oblique and normal incidence to the cutoff layer. The time evolution of backscattered signals, in excess of 30 dB, from seven distinct frequencies sampled simultaneously is presented across an L-to-H-mode confinement regime transition.

Magnetohydrodynamics in free surface liquid metal flow relevant to plasma-facing components

While flowing Liquid Metal (LM) Plasma-Facing Components (PFCs) represent a potentially transformative technology to enable long-pulse operation with high-power exhaust for fusion reactors, Magnetohydrodynamic (MHD) drag in the conducting LM will reduce the flow speed. Experiments have been completed in the linear open-channel LMX-U device [Hvasta et al 2018 Nucl. Fusion 58 01602] for validation of MHD drag calculations with either insulating or conducting walls, with codes similar to those used to design flowing LM PFCs for a Fusion Nuclear Science Facility [Kessel et al 2019 Fusion Sci. Technol. 75 886]. We observe that the average channel flow speed decreased with the use of conducting walls and the strength of the applied transverse magnetic field. The MHD drag from the retarding Lorentz force resulted in an increase of the LM depth in the channel that ‘piled up’ near the inlet, but not the outlet. As reproduced by OpenFOAM and ANSYS CFX calculations, the magnitude and characteristics of the pileup in the flow direction increased with the applied traverse magnetic field by up to 120%, as compared to the case without an applied magnetic field, corresponding to an average velocity reduction of ∼45%. Particle tracking measurements confirmed a predicted shear in the flow speed, with the surface velocity increasing by 300%, despite the 45% drop in the average bulk speed. The MHD effect makes the bulk flow laminarized but keeps surface waves aligned along the magnetic field lines due to the anisotropy of MHD drag. The 3D fringe field and high surface velocity generate ripples around the outlet region. It was also confirmed that the MHD drag strongly depends on the conductivity of the channel walls, magnetic field, and volumetric flow rate, in agreement with the simulations and a developed analytical model. These validated models are now available to begin to determine the conditions under which the ideal LM channel design of a constant flow speed and fluid depth could be attained.

3D ion gyro-orbit heat load predictions for NSTX-U

High power tokamaks operate with divertor heat loads capable of destroying the plasma facing components (PFCs). High fidelity heat load predictions are necessary to ascertain the PFC state for design and during operation. Typical heat flux calculations are 2D, time invariant, and assume that power flows directly along the magnetic field lines (the optical approximation). These assumptions neglect the complex 3D geometries employed to protect the PFCs, the time varying nature of the plasma and PFC thermal state, and the helical trajectories of ions with finite Larmor radii (the gyro-orbit approximation). An integrated software framework, the heat flux engineering analysis toolkit (HEAT), was developed to generate time varying optical heat loads applied to real engineering computer aided design (CAD) (Looby et al 2022 Fusion Sci. Technol. 78 10–27). Recently, an ion-gyro orbit module has been added to HEAT. This module calculates the helical trajectories of ions as they gyrate about the magnetic field lines using kinetic theory macro-particles to accelerate the calculation. First, the new gyro-orbit module will be presented. Next, a comparison to existing research is performed. Finally, an analysis of the gyro-orbit heat loads for NSTX-U is presented for diverted discharges using the engineering CAD models utilized for PFC fabrication. Including these gyro-orbit effects can enhance the PFC performance by ‘smearing’ out the magnetic shadows associated with the castellated fish-scaled geometry. Simultaneously, the helical trajectories can degrade performance when they load narrow regions on edges and corners with high heat fluxes. Analysis of the trade-offs between these competing effects is included, and regions for further investigation are identified.

Fluid, kinetic and hybrid approaches for neutral and trace ion edge transport modelling in fusion devices

Neutral gas physics and neutral interactions with the plasma are key aspects of edge plasma and divertor physics in a fusion reactor including the detachment phenomenon often seen as key to dealing with the power exhaust challenges. A full physics description of the neutral gas dynamics requires a 6D kinetic approach, potentially time dependent, where the details of the wall geometry play a substantial role, to the extent that, e.g., the subdivertor region has to be included. The Monte Carlo (MC) approach used for about 30 years in EIRENE (Reiter et al 2005 Fusion Sci. Technol. 47 172–86), is well suited to solve these types of complex problems. Indeed, the MC approach allows simulating the 6D kinetic equation without having to store the velocity distribution on a 6D grid, at the cost of introducing statistical noise. MC also provides very good flexibility in terms of geometry and atomic and molecular (A&M) processes. However, it becomes computationally extremely demanding in high-collisional regions (HCRs) as anticipated in ITER and DEMO. Parallelization on particles helps reducing the simulation wall clock time, but to provide speed-up in situations where single trajectories potentially involve a very large number of A&M events, it is important to derive a hierarchy of models in terms of accuracy and to clearly identify for what type of physics issues they provide reliable answers. It was demonstrated that advanced fluid neutral models are very accurate in HCRs, and at least an order of magnitude faster than fully kinetic simulations. Based on these fluid models, three hybrid fluid–kinetic approaches are introduced: a spatially hybrid technique, a micro–macro hybrid method, and an asymptotic-preserving MC scheme, to combine the efficiency of a fluid model with the accuracy of a kinetic description. In addition, A&M ions involved in the edge plasma chemistry can also be treated kinetically within the MC solver, opening the way for further hybridisation by enabling kinetic impurity ion transport calculations. This paper aims to give an overview of methods mentioned and suggests the most prospective combinations to be developed.

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.

Modelling of edge plasma dynamics with active wall boundary conditions

AbstractA self‐consistent 2D model is presented for transport in boundary plasma and plasma‐facing material walls. Plasma dynamics in the domain is represented by a 2D collisional plasma fluid model in the edge‐plasma code UEDGE (Rognlien et al., J. Nucl. Mater. 196–198 (1992) 347), and transport of hydrogen and heat in the wall is represented by a system of reaction–diffusion equations in the 1D wall code FACE (Smirnov et al., Fusion Sci. Technol. 71 (2017) 75). To account for variation of parameters along the wall, in the coupled model multiple instances of the FACE code run in parallel. The coupled model provides a tool for investigating a range of dynamic plasma–material interactions phenomena in 2D. For demonstration of its capability, one application of particular interest is the role of active wall in tokamak strike point sweeping proposed for mitigation of divertor heat loads. In the present study, the coupled calculations are applied to investigation of the impact of heat and hydrogen transport in the material wall on the divertor plasma and target heat load during sweeping of the target strike point for parameters of a high‐power tokamak.

Preliminary Cryogenic Layering by the Infrared Heating Method Modified with Cone Temperature Control for the Polystyrene Shell FIREX Target

The infrared (IR) heating method for a central ignition target with spherical symmetry is modified for the axisymmetric Fast Ignition Realization EXperiment (FIREX) target. The challenge is that the FIREX target pretends to be a thermally spherical shell. Our previous simulation studies (A. Iwamoto et al., Fusion Sci. Technol. 56, 427 (2009), A. Iwamoto et al., J. Phys.: Conf. Ser. 244, 032039 (2010)) have shown that the combination of volumetric heating in a fuel and cone temperature control has the potential to finish a uniform fuel layer. We have developed the IR heating system, dedicated to the FIREX target, with exclusive cone temperature control. The ability of solid fuel layering was examined by using an 826 µm polystyrene (PS) shell with a gold cone of 1.2 mm in length instead of the 500 µm FIREX target for easy observation. The system could control the profile of a solid fuel layer in the PS shell target. Eventually, the solid layer with the best sphericity of 92% was formed, and the RMS roughness of the inner surface was 44 - 49 µm in modes 1 to 100 and 14 - 26 µm in modes 5 to 100.

A Domestic Program for Liquid Metal PFC Research in Fusion

While high-Z solid plasma-facing components (PFCs) are the leading candidates for reactors, it is unclear that they can survive the intense plasma material interaction (PMI). Liquid metals (LM) PFCs offer potential solutions since they are not susceptible to the same type of damage, and can be “self-healing”. Following the Fusion Energy System Study on Liquid Metal Plasma Facing Components study that recently was completed by Kessel et al. (Fusion Sci Technnol 75:886, 2019) a domestic LM PFC design program has been initiated to develop reactor-relevant LM PFC concepts. This program seeks to evaluate LM PFC concepts for a Fusion Nuclear Science Facility (FNSF) or a Compact Pilot Plant via engineering design calculations, modeling of PMI and PFC components and laboratory experiments. The latter involves experiments in dedicated test stands and confinement devices and seeks to identify and answer open questions in LM PFC design. The new national LM PFC program is first investigating lithium as the plasma facing material for a flowing divertor PFC concept. Several flow speeds will be evaluated, ranging from ~ cm/s to m/s. The surface temperature will initially be held below the strongly evaporative limit in the first design; higher temperatures with strong evaporation will be considered in future concepts. Other topics of interest include: understanding of the hydrogen and helium interaction with the liquid lithium; single effect experiments on wetting, compatibility and embrittlement; and prototypical experiments for control and characterization of flowing LM. A path to plasma and future tokamak exposure of these concepts will be developed.

Inference of experimental radial impurity transport on Alcator C-Mod: Bayesian parameter estimation and model selection

We present a fully Bayesian approach for the inference of radial profiles of impurity transport coefficients and compare its results to neoclassical, gyrofluid and gyrokinetic modeling. Using nested sampling, the Bayesian impurity transport inference (BITE) framework can handle complex parameter spaces with multiple possible solutions, offering great advantages in interpretative power and reliability with respect to previously demonstrated methods. BITE employs a forward model based on the pySTRAHL package, built on the success of the well-known STRAHL code (Dux 2003 Fusion Sci Technol. 44 708–15), to simulate impurity transport in magnetically-confined plasmas. In this paper, we focus on calcium (Ca, Z = 20) laser blow-off injections into Alcator C-Mod plasmas. Multiple Ca atomic lines are diagnosed via high-resolution x-ray imaging crystal spectroscopy and vacuum ultra-violet measurements. We analyze a sawtoothing I-mode discharge for which neoclassical and turbulent (quasilinear and non-linear) predictions are also obtained. We find good agreement in diffusion across the entire radial extent, while turbulent convection and density profile peaking are estimated to be larger in experiment than suggested by theory. Efforts and challenges associated with the inference of experimental pedestal impurity transport are discussed.

Machine Learning Algorithms for Automated NIF Capsule Mandrel Selection

Small shells, approximately 2 mm in diameter, made from Poly(α-methylstyrene) (PAMS) are used as mandrels in the production of glow discharge polymer capsules located at the center of inertial confinement fusion experiments. The visual inspection process of microscope images of these shell mandrels, including detection of micron-sized defects on the shell surface as well as the handling and sorting, is a very labor-intensive, repetitive, and highly subjective process that stands to benefit greatly from automation. As part of an effort to decrease the number of labor hours spent in capsule handling, inspection, and metrology, the development of robotic systems was presented in a paper by Carlson et al., “Automation in Target Fabrication” [Fusion Sci. Technol., Vol. 70, p. 274 (2016)]. The current work expands the automated image acquisition systems developed previously and adds the use of convolutional neural networks to select capsules best suited for use in the downstream production process. Through the use of these machine learning algorithms, the selection process becomes robust, repeatable, and operator independent. As an added benefit the system developed as part of this work is able to provide defect statistics on entire shell batches and feed this information upstream to the production team.

Divertor for a steady-state gas-dynamic trap

An analysis is carried out into the possibility of using a device topologically equivalent to a nonparaxial magnetohydrodynamic (MHD) stabilizer as a divertor for the projected ALIANCE source of fusion neutrons based on the gas-dynamic trap, which should operate in a continuous mode. A side effect of adding a divertor to a linear trap is the expected improvement in plasma MHD stability, which was previously observed at the TARA (Casey et al 1988 Phys. Fluids 31 2009–16) and HIEI (Yasaka et al 2001 Fusion Technol. 39 350–3) facilities. To assess the effect of MHD stabilization by a divertor, the article indicates a method for finding stability rings, and also calculates the degree of expansion of the plasma flux in the divertor. The analysis showed that a good divertor, which provides a high degree of expansion, cannot be a good MHD stabilizer, which provides a large margin of stability. The divertor configurations made up of two and three coils are studied, their advantages and disadvantages are described. The final part of the article presents the calculations of the magnetic field in an end cell with a superconducting divertor designed for the GDMT gas-dynamic multi-core trap project (Beklemishev et al 2013 Fusion Sci. Technol. 63 46–51).

Electrodeposition of Graded Metal Coatings on Spherical Mandrels

Inertial confinement fusion energy research at the National Ignition Facility (NIF) requires the fabrication of hollow metal spherical capsules for the laser-driven fusion reaction. These sub-millimeter diameter capsules must have perfect coating uniformity and a specific gradient density, achieved by plating two metal elements with graded composition. We are investigating electrodeposition to produce hollow metal spheres as opposed to other deposition techniques (i.e. physical vapor deposition, film casting, etc.), because electrodeposition coats faster, more smoothly, with good thickness uniformity, and better grain size control. Electroplating metals on spheres presents its own specific challenges; the spherical samples must make contact with the electrified cathode, but must also remain in motion during plating to ensure an even coating. Additionally, electroplating on hollow spheres also involves the challenge of keeping the sphere submerged in the electrolyte. These issues were mitigated by enclosing the mandrel in an electrified cage which the electrolyte is flowed through, resulting in intermittent contact of the mandrel with the cathode and continuous flow of fresh electrolyte to the mandrel surface while plating. We have developed several approaches for electroplating alloys[1] and density gradients on metal spherical mandrels. We will present results from a parametric study that improved coating uniformity, obtained desired thickness, and achieved the required density gradients during electrodeposition. [1] C. Horwood, M. Stadermann, T.L. Bunn, Metal Alloy ICF Capsules Created by Electrodeposition, Fusion Sci. Technol. 73 (2018) 335–343. doi:10.1080/15361055.2017.1387458. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. IM Release# LLNL-ABS-748349

Investigation of Metal Ion Diffusion and Reduction Potential in Several Ionic Liquids

Electrodeposition is proving to be a promising technique for forming metal capsules and other components required for Inertial Confinement Fusion (ICF) experiments performed at the National Ignition Facility. While we have recently demonstrated that non-porous, defect free parts can be electrodeposited from a variety of water-based solutions,1,2 many of the preferred metals for the ICF program (beryllium, uranium, etc.) cannot be deposited because electrolysis of water occurs at the potential required to reduce such metals. To circumvent this, we are currently investigating Ionic Liquids (ILs) as solvents for metal electrodeposition, as the wide (~ 6 V) electrochemical window of ILs is expected to allow electrodeposition of Be, U, and many other metals.3 We are investigating the key parameters of metal ion diffusion and reduction potential for two model metals, silver and copper, in several different ionic liquids. We use electrochemical measurements along with first principles density functional theory (DFT) simulations to understand how ionic liquids affect diffusion and reduction potential, and this understanding will be applied to selecting appropriate ILs for deposition of metals such as Be and U. Funding was provided by Lawrence Livermore National Laboratory Directed Research and Development (LDRD) Grant 17-ERD-047. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. IM release LLNL-ABS-771850. (1) Horwood, C.; Stadermann, M.; Bunn, T. L. Fusion Sci. Technol. 2018, 73 (3). (2) Bunn, T.; Horwood, C.; Stadermann, M. Cathode System for Electrodeposition of Metals on Microspheres. US Patent 16336-000075-US-PS1, 2018. (3) Electrodeposition from ionic liquids, 2nd ed.; Endres, F., Abbott, A., MacFarlane, D., Eds.; Wiley-VCH: Weinheim, Germany, 2017.

ADDENDUM: About a New Fusion Reactor Scheme

A crucial problem of any steady-state fusion reactor is removal of the plasma exhaust. In this addendum to the paper “About a New Fusion Reactor Scheme” [Fusion Sci. Technol., Vol. 75, p. 197 (2019)], we analyze the possibility of achieving this goal in the fusion reactor scheme by using a model C stellarator type of divertor.