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In the original version of this Article, in the fifth paragraph of the section titled “Qneo for W7-X discharge 20180918.045,” errors appeared in reference to the configurations shown in the curves of Fig. 3a and associated Fig. 3b−e panels. In the sentence originally reading “This comparison is provided in Fig. 3a for the configurations (Fig. 3b) W7-X standard, (Fig. 3c) W7-X high-mirror, (Fig. 3d) LHD R0 = 3.6 m, and (Fig. 3e) LHD R0 = 3.75 m....,” the references to Fig. 3b−e should instead have referred to the lettered curves “b,c,d,e” shown in the graph of Fig. 3a. The text has now been amended to read: “This comparison is provided in Fig. 3a for the configurations (b) W7-X standard, (c) W7-X high-mirror, (d) LHD R0 = 3.6 m, and (e) LHD R0 = 3.75 m….” Additionally, a member of the W7-X Team was incorrectly listed as “M. Sanchez”; their name has been corrected to “E. Sánchez.” The original Article has been corrected online.

The European power system faces two main challenges: first, the increasing share of fluctuating energy sources in the power generation sector challenges the balance of generation and demand. Second, the expected shift towards using more electricity in the heating and transportation sectors leads to rising demand. To ensure both sufficient energy supply and long-term power system stability of the power system, fusion power is expected to play an important role in the future. This paper provides a sensitivity analysis of fusion power plant parameters in a power system with a large fraction of renewable energies. The results depend strongly on the underlying scenario assumptions, especially on future technical and economic improvements. Finally, the amount of fusion power investments depends on both demand level and relative costs of competing technologies.

The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

The neutronic assessment of a fusion power plant design is usually a challenging and time-consuming task involving experts from several disciplines in order to assemble the geometry, source, as well as carry out computationally heavy Monte Carlo transport simulations. In order to overcome this challenge, we present in this work a deterministic method, which combines all these aspects in a single framework that can directly calculate a key neutronics performance indicator, the neutron wall load (NWL), at minimal computational cost for arbitrary stellarator configurations. As our method is based on simple vector and matrix manipulations, a speed on the order of a few CPU-seconds is achieved, which makes it suitable for optimisation frameworks. We demonstrate with a simple optimisation algorithm that it is possible to use our method to generate a first wall with reduced ‘heterogeneity’ in the NWL distribution.

A first neutronics analysis of the Helical-Axis Advanced Stellarator (HELIAS) power reactor is conducted in this work. It is based on Monte Carlo (MC) particle transport simulations with the Direct Accelerated Geometry Monte Carlo (DAGMC) approach which enables particle tracking directly on the CAD geometry. A suitable geometry model of the HELIAS reactor is developed, including a rough model of a breeder blanket based on the Helium Cooled Pebble Bed (HCPB) breeder blanket concept. The resulting model allows to perform first neutronic calculations providing a 2D map of the neutron wall loading, a 3D distribution of the neutron flux, and a rough assessment of the tritium breeding capability. It is concluded that the applied methodology, making use of MC particle transport simulations based on the DAGMC approach, is suitable for performing nuclear analyses for the HELIAS power reactor.

We assess the magnetic field configuration in modern fusion devices by comparing experiments with the same heating power, between a stellarator and a heliotron. The key role of turbulence is evident in the optimized stellarator, while neoclassical processes largely determine the transport in the heliotron device. Gyrokinetic simulations elucidate the underlying mechanisms promoting stronger ion scale turbulence in the stellarator. Similar plasma performances in these experiments suggests that neoclassical and turbulent transport should both be optimized in next step reactor designs.

Isotope effects are one of the most important issues for predicting future reactor operations. Large helical device (LHD) is the presently working largest stellarator/helical device using super conducting helical coils. In LHD, deuterium experiments started in 2017. Extensive studies regarding isotope effects on transport have been carried out. In this paper, the results of isotope effect studies in LHD are reported. The systematic studies were performed adjusting operational parameters and nondimensional parameters. In L mode like normal confinement plasma, where internal and edge transport barriers are not formed, the scaling of global energy confinement time (τ E) with operational parameters shows positive mass dependence (M 0.27; where M is effective ion mass) in electron cyclotron heating plasma and no mass dependence (M 0.0) in neutral beam injection heating plasma. The non-negative ion mass dependence is anti-gyro-Bohm scaling. The role of the turbulence in isotope effects was also found by turbulence measurements and gyrokinetic simulation. Better accessibility to electron and ion internal transport barrier (ITB) plasma is found in deuterium (D) plasma than in hydrogen (H). Gyro kinetic non-linear simulation shows reduced ion heat flux due to the larger generation of zonal flow in deuterium plasma. Peaked carbon density profile plays a prominent role in reducing ion energy transport in ITB plasma. This is evident only in plasma with deuterium ions. New findings on the mixing and non-mixing states of D and H particle transports are reported. In the mixing state, ion particle diffusivities are higher than electron particle diffusivities and D and H ion density profiles are almost identical. In the non-mixing state, ion particle diffusivity is much lower than electron diffusivity. Deuterium and hydrogen ion profiles are clearly different. Different turbulence structures were found in the mixing and non-mixing states suggesting different turbulence modes play a role.

Advancing the current Advanced Driver Assistance Systems (ADAS) is coupled with introducing novel technologies into the automotive domain such as Light Detection and Ranging (LiDAR). LiDAR is attributed as a key-technology that will be one of the key enablers for safe and reliable automated driving. Considering the fact that vehicles nowadays rely on the driver in safety-critical situations leads to the problem that in a fully-automated driving scenario the vehicle needs to control every possible situation on its own. This increases the requirements and the overall safety level of the system but also for each component and needs a gradual transition from fail-safe to fail-operational behvior at least as long as the occupants and other road participants could be endangered.This publication introduces a novel system architecture of a fail-operational 3D Flash LiDAR System that enables dynamic system degradation during run-time as well as internal built-in self-test (BIST) for automated failure injection tests. The novel fail-operational system architecture is able to handle critical temperature ranges as well as long-term memory faults.

openaire: EC/H2020/633053/EU//EUROfusion Fast ion confinement is of major importance for the ignition of a burning fusion plasma. In future deuterium plasma campaigns of the Wendelstein 7-X stellarator, W7-X, the amount of triton burn-up is one possible measure for fast ion confinement. A well-established technique to observe triton burn-up is the 14 MeV neutron rate. In this paper, it is estimated whether an existing scintillating fibre neutron detector is also suited to measure triton burn-up in W7-X with sufficient accuracy. An estimation is presented, which can be applied to any tokamak or stellarator design and is one-dimensional in the minor radius. The inputs are profiles of density, temperature, and differential volume element as well as the triton slowing-down time. The estimation calculates the thermal deuteron fusion rate and the associated deuteron-triton fusion rate; thus, the triton burn-up generated 14 MeV neutron rate. It neither takes triton diffusion nor explicit losses into account. This thermally generated fusion rate is comparedto the neutral beam injection heating induced beam-plasma fusion rate.

Isotope effects of ECRH plasma in LHD were investigated in detail. A clear difference of transport and turbulence characteristics in H and D plasmas was found in the core region, with normalized radius ρ < 0.8 in high collisionality regime. On the other hand, differences of transport and turbulence were relatively small in low collisionality regime. Power balance analysis and neoclassical calculation showed a reduction of the anomalous contribution to electron and ion transport in D plasma compared with H plasma in the high collisionality regime. In core region, density modulation experiments also showed more reduced particle diffusion in D plasma than in H plasma, in the high collisionality regime. Ion scale turbulence was clearly reduced at ρ < 0.8 in high collisionality regime in D plasma compared with H plasma. The gyrokinetic linear analyses showed that the dominant instability ρ = 0.5 and 0.8 were ion temperature gradient mode (ITG). The linear growth rate of ITG was reduced in D plasma than in H plasma in high collisionality regime. This is due to the lower normalized ITG and density gradient. More hollowed density profile in D plasma is likely to be the key control parameter. Present analyses suggest that anomalous process play a role to make hollower density profiles in D plasma rather than neoclassical process. Electron scale turbulence were also investigated from the measurements and linear gyrokinetic simulations.

A first conceptual design of the Balance of Plant for a HELIAS power plant equipped with a Helium Cooled Pebble Bed (HCPB) Breeder Blanket (BB) is presented in this work. 5 inboard and 10 outboard cooling loops extract the heat from the blanket. Heat exchangers are used to transfer the thermal power from the helium to solar salt, which transports the heat out of the stellarator building and into an industrial 2-tank thermal energy storage system (IHTS/ESS). For electricity generation, the heat from the solar salt is transferred to the feedwater, which is boiled and superheated in the steam generator and then used in the turbine. In absence of a detailed design for the HCPB BB in stellarator geometry and the ex-vessel piping, a sensitivity study was carried out for the pressure drop of the cooling loops which showed that even small changes in the pressure drop have significant impact on the required pumping power. Finally, an exemplary runtime scenario is presented demonstrating the capability of a stellarator fusion power plant equipped with an IHTS/ESS to operate very flexibly (±20%) in an envisaged dynamic future energy system. Considering demand forecasts and the ESS, a stellarator fusion power plant (FPP) seems well equipped to cope with fluctuating grid demands within its technical limitations. If the conceptual idea for the HELIAS BoP as presented in this paper can be confirmed by more detailed technical studies, then a stellarator FPP could potentially contribute to grid stabilization in a future energy system while replacing fossil fuel power plants as a sustainable energy source.

Transport and confinement characteristics, and microinstabilities for high- and high- isotope plasmas in Large Helical Device are explored by using the gyrokinetic Vlasov simulation GKV with hydrogen isotope ions and real-mass kinetic electrons. The experimental data indicate that the thermal diffusivity is reduced in the deuterium-dominated plasmas, where the deviation from the gyro-Bohm scaling in the overall tendency and the strong anomaly of the electron heat transport are identified. Linear gyrokinetic analyses identify that the growth rates of the ion temperature gradient and trapped electron mode (TEM) instabilities in the deuterium plasmas are reduced due to the change in the profile gradients and/or the isotope ion mass, and the radial dependence of the mixing-length diffusivity is qualitatively consistent with the experimental tendency. Also, TEM-like turbulent fluctuations are examined by the using the phase contrast imaging measurement and the linear gyrokinetic calculation for the high- deuterium plasma.

Positive isotope effects have been found in electron cyclotron resonant heating plasma of the Large Helical Device (LHD). The global energy confinement time (τE) in deuterium (D) plasma is 16% better than in hydrogen (H) plasma for the same line-averaged density and absorption power. The power balance analyses showed a clear reduction in ion energy transport, while electron energy transport does not change dramatically. The global particle confinement time (τp) is degraded in D plasma; τp in D plasma is 20% worse than in H plasma for the same line-averaged density and absorption power. The difference in the density profile was not due to the neutral or impurity sources, but rather was due to the difference in the transport. Ion scale turbulence levels show isotope effects. The core turbulence (ρ = 0.5–0.8) level is higher in D plasma than in H plasma in the low collisionality regime and is lower in D plasma than in H plasma. The density gradient and collisionality play a role in the core turbulence level.

The deuterium (D) operation was initiated in the LHD in 2017. In the first campaign of the D experiments, we successfully extended the high temperature regime in the LHD. The new record of the ion temperature (Ti) of 10 keV associated with the ion internal transport barrier (ITB) was achieved due to several operational optimization. The thermal confinement characteristics of ITB plasmas were compared between hydrogen and D discharges. The effective ion thermal diffusivity of the ion-ITB plasmas was found to be smaller in the D discharges than that in the H discharges. The profiles of the Ti, the electron density, and the impurity of the high Ti plasmas strongly depended on the magnetic configuration and these profiles tended to peaked in the inward-shifted configuration. It was also found that the electron thermal confinement of the electron-ITB plasmas was clearly improved in the deuterium case. The GKV simulation showed the linear growth rate of TEM/ITG reduced in the plasmas with D both for the ion ITB and the electron ITB plasmas and qualitatively agreed with the tendency of the change in the thermal diffusivity obtained from the power balance analysis.

Three different approaches were investigated for the use of CAD models with Monte Carlo transport simulations of a simplified stellarator type verification geometry. These include the traditional Constructive Solid Geometry (CSG) approach with the geometry translated into Monte Carlo representation, the use of Unstructured Mesh (UM) geometry as available with MCNP6, and the direct particle tracking technique of DAGMC. The three methods were shown to be both applicable and suitable to represent the verification geometry and provide comparable results. With regards to the later application to a real stellarator geometry, the DAGMC and the UM approaches are preferred over the CSG approach due to the inherent limitations of the latter.

Transport studies of pellet fuelling experiments on LHD are reported. Spatio-temporal evolutions after pellet injection into LHD discharges show cases where central density increases on the time scale of particle transport processes. Both the temperature gradient and the density gradient change during the density relaxation, the latter even in sign. The resulting thermodynamic forces influence radial electric fields - both as a driving term but also by, e.g. affecting the E r dependence of ion transport. Magnetic fluctuations have been found to be induced by pellet injection but die out with the relaxation of the pressure profile.

An indirect effect of the electron cyclotron current drive (ECCD) on the ion temperature in the plasma core region was observed in the Large Helical Device. The reference (no ECCD) discharge with a central ion temperature Ti0 of ∼3.0 keV is operated by a standard high ion temperature discharge procedure. To investigate the ECCD effect, a co- or counter-ECCD was applied to the reference discharge, and was turned off immediately before the Ti0 peaked in the reference discharge. In the co-ECCD and counter-ECCD applications, the Ti0 temporarily increased and decreased by ∼0.5 keV from Ti0 in the reference discharge, respectively. The mechanism of this phenomenon is presently unclear, but may be exploited as a practical knob for controlling the central ion temperature.

In the stellarator community, a broad variety and diversity of three-dimensional magnetic configurations exist. In this chapter, the respective power plant studies of the major helical magnetic confinement concepts are presented, namely for the heliotron, the advanced stellarator and the compact stellarator concept. Since the diversity of the magnetic configurations is also reflected in the proposed technological solutions, the individual engineering concepts (in particular for blanket and divertor) are presented and brought in context by their advantages and drawbacks. Further, an outlook is given regarding the current developments and future prospects of stellarator fusion power plants.

The superconducting stellarator Wendelstein 7-X is currently being commissioned. First plasmas are expected for the second half of 2015. W7-X is designed to overcome the main drawbacks of the stellarator concept and simultaneously demonstrate its intrinsic advantages relative to the tokamak—i.e., steady-state operation without the requirement of current drive or stability control. An elaborate optimization procedure was used to avoid excessive neoclassical transport losses at high plasma temperature, while simultaneously achieving satisfactory equilibrium and stability properties at high $\beta $ in combination with a viable divertor concept. In addition, fast-ion confinement must be consistent with the requirements of alpha-heating in a power plant. Plasma operation of Wendelstein 7-X follows a staged approach following the successive completion of the in-vessel components. The main objective of Wendelstein 7-X is the demonstration of steady-state plasma at fusion relevant plasma parameters. Wendelstein 7-X will address major questions for the extrapolation of the concept to a power plant. These include divertor operation at high densities, plasma fueling at high central temperatures, avoiding impurity accumulation, and an assessment of the effect of neoclassical optimization on turbulent transport and fast-ion confinement. A power plant concept based on an extrapolation from Wendelstein 7-X, the helical advanced stellarator, has been developed.

System codes are commonly employed for the analysis and conceptual design of fusion reactors. For the helical-axis advanced stellarator (HELIAS) line, a new set of system code models have been developed to account for the stellarator-specific 3-D aspects. The models have recently been implemented in the systems code PROCESS and verified with respect to different test cases. After having established confidence in the stellarator models, system studies were carried out for the five-field-period HELIAS case to define the accessible reactor design window. In the multidimensional physics and engineering parameter space, sensitivity studies are carried out for the reactor regime to ascertain tradeoffs between different parameters and costs. Exemplary design points are analyzed in more detail using the plasma operation contour approach, which, for example, can be used to determine the optimum start-up path to ignition. Finally, with a common set of nondevice-specific models, the PROCESS framework allows a direct comparison of tokamaks and stellarators. Although the five-period HELIAS is a larger machine in terms of major radius, the required mass for both concepts is comparable leading to similar construction costs.

Systems codes are commonly employed for the analysis and conceptual design of next-step burning-plasma devices. For the helical-axis advanced stellarator (HELIAS) line a new set of systems code models have been developed to account for the stellarator-specific 3D aspects. The models have recently been implemented in the systems code PROCESS and verified with respect to different test cases.

In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

In order to study design points of next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. The code package PROCESS is such a tool, widely used for tokamak systems studies. In this work, the implementation and verification of a HELIAS module into PROCESS is addressed. These HELIAS models include: a plasma geometry model based on Fourier coefficients, a basic island divertor model, as well as a coil model which combines scaling aspects based on the Helias 5-B reactor design in combination with analytic inductance and field calculations. The models are verified firstly with respect to W7-X. Secondly, the generality of the models is used to represent the tokamak which is compared against the original tokamak PROCESS models using a DEMO design as reference case. Both approaches show very good agreement.

The optimized stellarator Wendelstein 7-X (W7-X) is designed to have an approximately quasi-isodynamic (QI) magnetic configuration with reduced neoclassical transport compared to a classical stellarator, and turbulent transport is expected to be a significant source of heat transport across the plasma minor radius. The Ion Temperature Gradient (ITG) and Trapped Electron Mode (TEM) are the dominant turbulence channels in low beta W7-X plasmas, and gyrokinetic modelling indicates two major differences between ITG/TEMs in W7-X and previous tokamak and stellarator experiments. The ITG mode is predicted to be highly localized along a helical band in W7-X, causing a reduction of the associated turbulence, and the nearly QI configuration of W7-X is predicted to be resilient to collisionless TEMs. In this contribution, the electron heat transport is compared to neoclassical predictions, and the stiffness in the electron heat transport is compared to TEM/ITG mode driven turbulence in two magnetic configurations of W7-X: one nearly QI configuration as well as a configuration with deliberately reduced QI. Additionally, electron temperature and plasma density fluctuations are compared to linear and nonlinear gyrokinetic calculations to assess the stabilization of TEM turbulence in W7-X.

With a lack of plasma disruptions and current-driven instabilities, stellarators are potentially an attractive option for a fusion power plant. Previous system studies have been performed to optimise a HELIAS (HELIcal-axis Advanced Stellarator) 5-B power plant using the systems code process , however these have been based around a single design point. In reality there is a lot of uncertainty extrapolating from present day devices and understanding. In this paper we study how this will affect the design by identifying eight uncertainty distributions on the input. We then perform parameter studies and Monte-Carlo based analysis to look at the impact on fusion power and divertor heat load. We find that the two uncertainties that have the largest impact on the fusion power are the helium primary coolant mechanical pumping power and the energy multiplication in the blanket and shield. Eighty-three per cent of our solutions are within a tolerable divertor heat load, however this is additionally influenced by the tungsten impurity levels. In order to stay below the density cut-off limit for Electron Cyclotron Resonance Heating, the confinement time needs to be enhanced relative to the ISS04 scaling relation to produce acceptable performance. By identifying the highest impact design parameters, we are able to highlight that research into the blankets should be prioritised to reduce overall design uncertainty.

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 × 1019 m−3 , central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre. EURATOM 633053

This paper reports on the effect of on- and off-axis heating power deposition on the impurity confinement in purely electron cyclotron resonance heated He plasmas on the stellarator Wendelstein 7-X. Therefore, impurity transport times τ I have been determined after Fe impurity injections by laser ablations and monitoring the temporal impurity emissivities by the x-ray imaging spectrometer HR-XIS. A significant increase of τ I has been observed when changing the power deposition from on- to off-axis heating with energy confinement times τ E being mainly unaffected. In addition, the scaling of impurity transport properties with respect to a variation of heating power P ECRH and electron density n e has been investigated by keeping the heating power deposition on-axis. The observed τ I scaling compares well to known τ I scaling laws observed in other machines. A comparison of τ I and τ E yields an averaged ratio of τ E /τ I = 1.3 and transport times in the range of τ I = 40–130 ms and τ E = 40–190 ms. Comparing those absolute values to neoclassical predictions supports the recently observed nature of anomalous transport in Wendelstein 7-X, given within the up to now investigated operational parameters.

Due to foreseen improvement of neoclassical confinement in optimised stellarators - like the newly commissioned Wendelstein 7-X (W7-X) experiment in Greifswald, Germany - it is expected that turbulence will significantly contribute to the heat and particle transport, thus posing a limit to the performance of such devices. In order to develop discharge scenarios, it is thus necessary to develop a model which could reliably capture the basic characteristics of turbulence and try to predict the levels thereof. The outcome will not only be affordable, using only a fraction of the computational cost which is normally required for repetitive direct turbulence simulations, but would also highlight important physics. In this model, we seek to describe the ion heat flux caused by ion temperature gradient (ITG) micro-turbulence, which, in certain heating scenarios, can be a strong source of free energy. With the aid of a relatively small number of state-of-the-art nonlinear gyrokinetic simulations, an initial critical gradient model (CGM) is devised, with the aim to replace an empirical model, stemming from observations in prior stellarator experiments. The novel CGM, in its present form, encapsulates all available knowledge about ion-driven 3D turbulence to date, also allowing for further important extensions, towards an accurate interpretation and prediction of the 'anomalous' transport. The CGM depends on the stiffness of the ITG turbulence scaling in W7-X, and implicitly includes the nonlinear zonal flow response. It is shown that the CGM is suitable for a 1D framework turbulence modeling.

As a starting point for a more in-depth discussion of a research strategy leading from Wendelstein 7-X to a HELIAS power plant, the respective steps in physics and engineering are considered from different vantage points. The first approach discusses the direct extrapolation of selected physics and engineering parameters. This is followed by an examination of advancing the understanding of stellarator optimisation. Finally, combining a dimensionless parameter approach with an empirical energy confinement time scaling, the necessary development steps are highlighted. From this analysis it is concluded that an intermediate-step burning-plasma stellarator is the most prudent approach to bridge the gap between W7-X and a HELIAS power plant. Using a systems code approach in combination with transport simulations, a range of possible conceptual designs is analysed. This range is exemplified by two bounding cases, a fast-track, cost-efficient device with low magnetic field and without a blanket and a device similar to a demonstration power plant with blanket and net electricity power production.