Initial Exploration of HighField Pulsed Stellarator Approach to Ignition Experiments
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Abstract
In the framework of fusion energy research based on magnetic confinement, pulsed highfield tokamaks such as Alcator and FTU have made significant scientific contributions, while several others have been designed to reach ignition, but not built yet (IGNITOR, FIRE). Equivalent stellarator concepts, however, have barely been explored. The present study aims at filling this gap by: (1) performing an initial exploration of parameters relevant to ignition and of the difficulties for a highfield stellarator approach, and, (2) proposing a preliminary highfield stellarator concept for physics studies of burning plasmas and, possibly, ignition. To minimize costs, the device is pulsed, adopts resistive coils and has no blankets. Scaling laws are used to estimate the minimum field needed for ignition, fusion power and other plasma parameters. Analytical expressions and finiteelement calculations are used to estimate approximate heat loads on the divertors, coil power consumption, and mechanical stresses as functions of the plasma volume, under wideranging parameters. Based on these studies, and on assumptions on the enhancementfactor of the energy confinement time and the achievable plasma beta, it is estimated that a stellarator of magnetic field B ~ 10 T and 30 m^{3} plasma volume could approach or reach ignition, without encountering unsurmountable thermal or mechanical difficulties. The preliminary conceptual device is characterised by massive copper coils of variable crosssection, detachable periods, and a lithium wall and divertor.
Keywords
Stellarator Ignition parameters Resistive magnets Monolithic supportIntroduction
Fusion energy is widely considered a potentially clean and abundant energy source [1, 2]. Current mainline research in magnetic confinement fusion is based on the tokamak concept [3], in spite of the important drawback posed by the possibility of disruptions and the challenge of steadystate operation. Correspondingly, alternatives based on the stellarator concept have also been developed [4, 5, 6]. Among them are a design of an ignition experiment (HSR4/18i [7]) and a burningplasma stellarator concept [8]. Many tokamaks and stellarators were built and operated to investigate a variety of fusion plasma problems [9, 10, 11]. However, understanding the physics of burning plasmas remains a research challenge [12, 13].
For both concepts, tokamaks and stellarators, a higher magnetic field leads to a smaller and potentially more costeffective experimental device [14, 15]. Additionally, devices equipped with resistive magnets, of moderate cost, are suited to produce pulses of few seconds (longer or much longer than the energy confinement time and alphaparticle slowing down time), which are appropriate to perform a diversity of burning plasma experiments.
In tokamaks, several high magnetic field devices have been satisfactorily built and operated to explore and validate this approach, e.g. Alcator and FTU [16, 17]. Other highfield experimental tokamaks have been designed to reach ignition but not built yet, e.g. IGNITOR and FIRE tokamaks [18, 19]. The IGNITOR design employs massive cryocooled copper magnets and pursues plasma ignition using a high magnetic field B ~ 13 T in a small plasma volume V ~ 10 m^{3}, at β ~ 1.2% (β is the plasma kinetic pressure normalized to the magnetic pressure). Similarly, FIRE is another highfield tokamak design (B ~ 10 T, V ~ 20 m^{3}) aimed at approaching ignition which also uses cryocooled copper magnets.
However, the exploration of highfields in stellarators has been scarce. One exception is FFHR2 [20], but that is a power plant design, not an experimental device. Consequently, a stellaratorbased, highfield, high power density and resistivemagnet approach to the production of plasma ignition experiments appears fundamental. It would shed light, rapidly and at modest cost, on essential reactorrelevant physics and technology, and thus, it deserves exploration.
In this context, the present paper proposes a highfield stellarator path toward the study of burning plasmas. As an initial approximation, the work: (1) explores the essential physics and technological parameters of ignitioncapable experimental stellarators, particularly the operational limits and difficulties at high fields, and (2) derives an initial stellarator conceptual design.
The parameter scan is deliberately broad to provide rough initial estimates of possible operating points for the design. Firstly, we estimate the minimum magnetic field needed for ignition and the fusion power as a function of the confinement enhancement factor h_{E} (as in International Stellarator Scaling 2004, ISS04 [21]), β and V. Subsequently, we study the technological parameters: heat load on the divertor targets, electric power needed to feed the resistive magnets and stresses on the coil supports, also as a function of h_{E}, β and V. Among the potential operating points, a reasonable one is downselected at the frontier of the physics and technological limits. Finally, from the operating point and the studies performed, the definition of a possible highfield ignitioncapable experimental stellarator is presented, called i–ASTER. This is characterised by massive copper coils of variable crosssection (so as to reach high fields with feasible power supplies), a lithium divertorwall to try to deal with the high power density, and absence of blankets to lower costs.
The work is organized as follows. In “Assumptions and Governing Equations: Ignition Condition” section we formulate the governing physics equations. The technological parameters and constraints are presented in the next sections: heat load on the divertors (“Power Load on Divertor Targets” section), power needed to operate the resistive magnets (“Power Dissipated in Resistive Magnets” section) and stresses in the coil support structure (“Estimation of Stress in Coil Structures” section). Finally, the resulting specifications of a possible ignition stellarator concept are presented in “Definition of iASTER”.
Assumptions and Governing Equations: Ignition Condition
A power balance equation and a scaling law for the energy confinement time are the essential physics equations involved. Additionally, the fusion power generated under ignition or the maximum possible plasma density, equal to the Sudo density limit [22], could have been minimized. Instead, we decided to minimize the magnetic field since it clearly correlates with the cost of the coils and their support structures [23]. Only an initial estimate of possible operating points is sought here. Detailed plasma calculations using advanced codes [24, 25] are left for future work, as the design advances.
The governing physics equations assume a scalable device for scanning the plasma volume and the device size. Thus, all proportions and all shapes (e.g. of the coils and their support structures) are preserved, and all dimensions, such as the distance from plasma edge to the winding surface, scale with a scaling factor.
Under such premises, two rather extreme values of h_{E} (0.75, 1.5) are considered in the remainder, as well as three values of the volumeaveraged beta limit < β > _{lim} (2.5%, 5%, and 10%). Values inbetween these limits are conceivable and thus, potential operating points. These limits were selected as follows, according to experimental and theoretical data.
An enhancement factor h_{E} around 1.5 was experimentally achieved in some highβ pulses in W7AS and slightly lower in the LHD inwardshifted configuration [21]. Calculations have predicted h_{E} ~ 2 for W7X [24], but this is yet to be proven experimentally.
Experimentally, W7AS achieved a maximum < β > = 3.2% [11] and < β > = 5% was demonstrated in LHD [26]. Up to β_{lim} ~ 7% may be achievable for the low aspect ratio A ~ 4.5 NCSX [27, 28]. β_{lim} ~ 10% is calculated for a large aspect ratio A ~ 10 quasihelical stellarator [29], and slightly lower β_{lim} ~ 8.5% for A ~ 12 in QIP6 (QuasiIsodynamic with poloidally closed contours of constant B of 6 periods) [30, 31]. Second stability regimes of high beta 7–20% in compact stellarators have been theoretically predicted [32, 33] but are yet to be experimentally proven.
Power Balance
Energy Confinement
Different scaling laws are available in the literature [38], with different coefficients C_{0} and different exponents, but here we follow the ISS04 international stellarator scaling [21]. Here R is the plasma major radius, P the effective heating power (≡ P_{α_heat}) and ι_{2/3} the rotational transform at r = 2/3a, where a is the plasma minor radius.
An aspect ratio A = 6 is assumed, as a rough average between A ~ 4.5 in ARIESCS [39], A ~ 6 in HSR3/15 (Helias Stellarator Reactor of 3 periods) [40] and A ~ 7 in QIP3 (QuasiIsodynamic stellarator with poloidally closed contours of 3 periods) [30].
Additional assumptions include ‘intermediate’ temperature and density profiles, similar to HSR4/18i [7]—that is, neither too flat, nor too peaked. Flatter profiles would yield higher fusion power but require higher B for ignition. More peaked profiles have been obtained in stellarators [41] but it is unknown whether they would be feasible in burning plasmas.
Estimate of Minimum B for Ignition
Density and Temperature Needed for Ignition, Fusion Power
The ignition temperature is independent of h_{E,} β_{lim} and V. For the assumed Z_{eff}, pressure profile and A, the central temperature evaluates to T_{0.ig} = 14.6 keV.
Power load on Divertor Targets
It is assumed that the incident power equals the alpha heating power P_{α}, in the limit of negligible power radiated by the divertor mantle and SOL.
Wetted area and concentration factor K_{d} in different fusion devices
As known, divertorrelated challenges could limit the attractiveness of fusion as a competitive energy source, both in stellarators and tokamaks [47, 48]. Divertors are less critical in shortpulse physics experiments, but still plasma purity and thermal shocks on the walls and divertor targets are relevant.
 1.
It is assumed that a reasonable increment of 50% of wet area relative to W7X divertor (increase from 2 to 3 m^{2} in Table 1) is possible by modern optimization, resulting in K_{d} ~ 40.
 2.Sweeping of the divertor legs on the targets by slightly changing the currents in coils. It would change the size and position of the magnetic islands [11, 44], increasing the wet area and smoothing the heat load on the targets [49, 50, 51]. Doubling the wet area of an improved quasiisodynamic configuration is assumed in Fig. 4, K_{d} ~ 20.
 3.
50% of the power is radiated by the plasma edge, also considered in Fig. 4.
The resulting heat loads are plotted in Fig. 4.
If such conditions are not met, it can be shown that ignition could be achieved by reducing β to ~ 2.5% or less and increasing B. This, however, would largely reduce the attractiveness of the approach, unless a solution is adopted—probably based on liquid lithium, which may withstand high P_{d}. As an added benefit, low recycling Li walls enhanced confinement in TFTR [52], TJII [53], NSTX [54] and other devices [55, 56] by various amounts, ranging between 25% and 100%. Liquid lithium does not erode or blister. Low Li impurity in the core plasma was obtained in NSTX and TFTR [55], which allowed low Z_{eff} (~ 1.3), e.g. in TFTR [37]. Drawbacks of lithium utilization, like oxidation, fire risk, tritium retention and others are cited in Ref. [57].

Jets of liquid metal droplets flowing on limiters or divertors. As an example, GaInSn droplets of 2–4 mm diameter and 25 m/s extracted 510 MW/m^{2} from the T–3 M tokamak [58, 59].

Liquid Li limiters or walls based on a Capillary Porous System (CPS), as tested in FTU [60, 61] and TJII [62]. In FTU they withstood an average of 2 MW/m^{2} and brief (300 ms) peak values of 5 MW/m^{2} (see Fig. 12 in Ref. [61]). Experiments with a CPS liquid lithium limiter on T11 M tokamak [63] achieved 10 MW/m^{2} on the limiter (0.3 s pulse), and 30 MW/m^{2} including radiation from Li ions.

Beams of high speed (> 100 m/s) Li droplets [64], as theoretically proposed for the ITER divertor.

Molten (tin) shower jets are theorized for the FFHR reactor [65].
Indeed, promising high power extraction systems could be properly tested and enhanced in the present high power density approach.
The average neutron wall load (Fig. 4) is calculated as the total neutron power divided by the plasma surface.
Power Dissipated in Resistive Magnets
The effective cross section of the coils is maximized in order to reduce the coil resistance and lower the Ohmic power dissipated in the coils, as in Refs. [35, 66] and in Fig. 7. As a result of this design, each coil presents variable crosssection in poloidal direction. The cross sections tend to be smaller on the inboard of the stellarator and larger on the outboard (Fig. 7), leading respectively to a local increase and local reduction of dissipated power, partially compensating each other.
Ports are not defined in this initial model for electric calculations, but they will be small as explained in “Resistive Magnets” section and would not hinder the massive quasicontinuous coils.
A simple analytical expression is derived in “Analytic Approximation to Dissipated Power” section for the power dissipated in the coils. Some factors involved in that expression are computed in “Finite Elements Results” section with the aid of finite elements.
Analytic Approximation to Dissipated Power
Finite Elements Results
Power dissipation is calculated by finite elements in the CASTELL code, using the configuration depicted in Fig. 7 except that we treat the trapezoidal crosssections of that figure as rectangular. In addition, values ξ = 1.75, ε = 0.5 are used for QIP3, ξ = 2, ε = 1 for HSR3, and f_{i} = 6/7 for both. It results that the analytical expression (9) agrees, with deviations lower than 20%, with the timeconsuming finite elements calculation for QIP3 and HSR3, taking a fixed f_{s} = 1.3 (in comparison, for a tokamak f_{s} = 1). From the study performed for QIP3 and HSR3 configurations, 1.2 < f_{s} < 1.4 is expected for typical stellarator magnetic configurations.
Current Density and Coil Temperature
The current density j_{s} in the coils is evaluated at the crosssections S located at the major radius R (Fig. 5) and averaged over all coils. Nevertheless, the current density is higher in certain locations. We denote by f_{c} the concentration factor for the maximum current density relative to j_{s} (j_{max} = f_{c} j_{s}). f_{c} is calculated by finite elements in CASTELL code as the ratio of the average cross section of all the finite elements for all the coils to the minimum cross section found among the coils. As an example, f_{c} = 5 for QIP3 and f_{c} = 6 for HSR3 was calculated for the conditions in “Finite Elements Results” section f_{c} < ~ 6 is expected for nonquasiisodynamic stellarators since there is not a mirrorlike magnetic field.
Being, t the pulse length (5 τ_{E}), C_{p} the volumespecific heat of the material, P_{coils} total power dissipated in coils from Eq. (9), V_{totCu} total volume of copper in all coils, and the remainder as in “Power Dissipated in Resistive Magnets” section.
Limitations and Discussion
The large thickness of the magnets for reasonable power supplies is a concern. Thickness as wide as the plasma minor radius (ε = 1) is taken for Fig. 6. Despite that, additive manufacturing can help the fabrication of such thick layer(s) of conductor and insulation, as being investigated for stellarator coils in Refs. [69, 70, 71].
Also, the fabrication method for the variable crosssection coils requires future exploration. Water jet cutting of copper sheets and winding of the resulting conductors in additively manufactured grooves is a construction option. Another alternative is the use of a single, properly grooved thick metal layer conformal to the vacuum vessel, with insulating layers in the grooves, similarly to the concept depicted in Ref. [66].
The massive resistive coils of variable crosssection involve new calculation methodologies and advanced magnetic error prediction. The coil width, number of coils and the number of layers per coil has to be decided according to: i) finite element analysis of the current paths in the wide coils, and ii) the nonuniform increase of copper temperature and thus differential increase of resistivity due to Joule heating. Such advanced calculations will be investigated in next development phases.
Estimation of Stress in Coil Structures
The yield tensile strength of the coil support materials and insulation constrain the maximum achievable B.
In this section, first an analytic approximation is deduced and then a specific finite elements calculation is performed.
Analytic Approximation of Stress
Let us approximate the stellarator coils as if they were circular and uniformly distributed, in the toroidal direction, in a monolithic support of thickness d = ψ a .
The maximum stress in the structure is σ _{max} = f_{σ} σ_{s,} where f_{σ} is a stress concentration factor. Finite element calculations presented in the next section will show that f_{σ} ~ 2 – 3, depending on the type of stellarator.
Finite Element Calculation
Loads due to the weight of the structure are not considered, and openings through the structure are neglected. The central ring is modelled as a thin hexagon in order to avoid impacting the calculation.
To fix the ideas, we set V = 30 m^{3}, B = 9.8 T (see the h_{E} = 1.5, β_{lim} = 5% case in Fig. 1), ψ =0.5 and a current of 1.6 MA in each coil. Under these conditions, 3600 elements of force on 144 coils were calculated by the CASTELL code, introduced in the Finite Element Analysis (FEA) module of CATIA and, applied on the support structure. This model hinders the calculation of the stress in the coils and intercoil insulation.
The maximum stress in the monolithic support (σ _{max} ~ 600 MPa) is located at the inboard of the curved section. Such value is 2.5 times higher than the result (σ _{s} = 245 MPa) from Eq. (12), thus f_{σ} = 2.5.
Limitations and Discussion
This initial stress calculation does not tackle the insulation stress, which remains for future detailed studies. High strength insulation might be required.
The type of magnetic configuration changes the location of the areas of maximum stress, i.e. [72], but the approach of considering an averaged value σ_{s} and an stress concentration factor f_{σ} is still helpful.
Local adjustment or optimization of the thickness of the structure could smooth stress and deformation on the full structure.
In comparison to tokamaks, the larger aspect ratio of stellarators decreases the forces in the inboard of the torus [15] but the stress concentration factor in stellarators is unfavourable. In spite of this, the maximum stress in the monolithic support in iASTER resulted in similar levels to the maximum stress in the coil support of a high field tokamak like IGNITOR, ~ 500 MPa, [18].
Definition of iASTER
i–ASTER is a highfield, small size and resistivemagnet stellarator concept designed to reach ignition and study burning plasmas. It is not a power plant prototype.
Mission and General Characteristics
iASTER aims at, rapidly and at modest cost, achieving and understanding ignition, and studying alphaparticle physics in ignited or nearignited plasmas in a small fusion device. This physics will be only partially investigated in ITER. Thanks to its high powerdensity, i–ASTER could serve the additional goal of testing and optimizing power extraction systems (e.g. lithiumbased) and studying the plasmawall interaction. Indirectly, it would complement the stellarator research line in the high plasma pressure range, advance technologies for high field fusion devices and for the manufacturing of strong stellarator magnets.
Pulses are foreseen to last few seconds (much longer than the energy confinement, alphaparticle slowing down time and other timescales of interest) and to be repeated with a low dutycycle (~ 1000 pulses during a ~ 10 year lifetime). This approach reduces cost and neutronic issues and still accomplishes the research mission stated above. The dutycycle is selected as an initial conservative value from estimations on neutronics effects (i.e. on copper resistivity) and, to achieve undemanding and slow cooling of coils between pulses. The model to perform such estimations is an ignitioncapable stellarator working at the frontier of the physics and technological limits (minimum size device) whose size is independent of the duty cycle. The optimization of the device size based on the ratio of number of pulses to facility cost is out of scope of the present work.
In the spirit of reducing costs, and compatible with short pulses, iASTER adopts resistive magnets, which are faster to manufacture and simpler to operate than superconducting coils. Also, resistive magnets allow faster tests, avoid cryostat, cryoplant and cooldown time, allocate extra space for the plasma due to thinner shielding, simplify radioactive waste recycling and, thus, moderate costs.
Main Design Features of iASTER
The three essential technological characteristics of i–ASTER (massive resistive magnets, detachable periods and Li divertorswalls) are described in the three subsections below. Subsequently, four complementary features are mentioned.
Resistive Magnets
The external surface of the torus would be covered by a thick layer or multilayer of copper, forming a series of wide modular coils of variable cross section (Figs. 7 and 10). The magnets would work adiabatically and a minimal cooling system would remove the heat during the long time between consecutive pulses. Aluminium is a backup alternative to copper.
Detachable (Half)Periods
The periods or halfperiods of the stellarator shall be easily separated from adjacent periods for easy assembly and maintenance. A (half)period would be removed from the torus and immediately, a refurbished or new one would be installed in order to minimize the maintenance downtime, e.g. coil replacement, which will be critical in the future power plants. Detachable periods were previously studied for superconducting coils [75] and appear equally advantageous and easier to realize for resistive magnets. The accuracy of the reassembly is a concern, but appropriate remote maintenance techniques are highly accurate [76, 77]. For example, a circular central ring (Fig. 12) would facilitate accurate reassembly. Larger twisted modular coils located at the vacuum vessel interfaces would facilitate (dis)assembly and port allocation (Fig. 12). Large modular coils were also planned in certain versions of NCSX stellarator [78].
Lithium DivertorWall
An island divertor [11, 79] and a firstwall almost entirely covered with lowtemperature (low recycling) liquid lithium is planned for iASTER. The latter could be realized by electrostatic/centrifugal spraying or by evaporation [80] of lithium on a thin Capillary Porous System (CPS) mesh (~ 0.2 mm thickness), similarly to the approach in Ref. [62]. The mesh is locally heated during coating from inside the vacuum vessel for proper Li deposition in the capillary mesh. The CPS is located on a thick copper substrate (the first wall) coated with a thin protective film of a Li compatible material (W or Mo). The lithium in the CPS is solid before the plasma discharge, at room temperature or slightly higher, and it is liquefied after the pulse start. For simplicity, heaters [62] are not planned in the copper substrate. The copper substrate at the divertor target areas would reach surface temperature 1200–1300 °C (for 30 MW/m^{2} thermal load and 2 s pulse), which would melt Cu and volatilize Li. Dry (tungsten or CFC) divertor targets enduring ~ 30 MW/m^{2} heat load [81, 82] or, advanced Libased systems (jets of droplets, beams of droplets or shower jets, “Power Load on Divertor Targets” section) to dissipate a fraction of the heat load before reaching the LiCPS, would allow withstanding the intense heat load.
Pulse Length
Ignition conditions are to be maintained for few energy confinement times τ_{E} (5 τ_{E} assumed here, comparable to 10 τ_{E} in FIRE [19]). The discharge is approximately 40 times longer than the alphaparticle slowing down time [83], thus enabling the study of alpha particles and their confinement.
Distance from Plasma to Coils
The copper coils are as thick (ε = 1) and as far from the Last Closed Flux Surface (LCFS) (ξ = 2) as reasonably possible for a smooth plasma shape of the HSR3 type.
For ξ = 2, A = 6 and V = 30 m^{3}, this gives Δ’ = 0.3 m.
No space is allocated for the breeding blankets in i–ASTER because breeding Tritium goes beyond the scope of the device. Besides, Δ’ is too small to accommodate a breeding blanket.
Heating System
The heating systems would only be used to ignite the plasma. The frequency needed for ECRH heating at B = 9.8 T, even at first harmonic, is unusually high (275 GHz), which will increase the cost of the gyrotrons. The cutoff density for Omode ECRH is 9.2 × 10^{20} m^{−3}, slightly lower than required (Fig. 2). This implies that the plasma will be slightly overdense and will require the excitation of Electron Bernstein Waves by means of OrdinaryeXtraordinaryBernstein mode conversion—a technique wellestablished in the W7AS stellarator and elsewhere [84].
Essential Diagnostics Strategy
Detailed integration of plasma physics (e.g. magnetic configuration, experimental plan) and technology (e.g. coil design, access for diagnostics) shall be produced. In the current initial design, two main ports (Fig. 12) are considered available for diagnostics (“Resistive Magnets” section), which will be complemented with some small ports. The diagnostics shall be designed and accommodated in each port in a fully integrated manner, for miniaturization. In a first stage, the diagnostics would be committed to plasma operation and machine protection (characterization of density and temperature profiles, neutron diagnostics, monitoring Li divertorwall conditions, and the few plasma control diagnostics needed in a stellarator). In a 2^{nd} stage, they would be mostly dedicated to study energetic particle dynamics (e.g. alphaparticle induced instabilities, alphaparticle losses and confinement). The FIRE tokamak diagnostics [85] are a reference for iASTER.
Size and Materials for iASTER.v1 According to Limits
Values of h_{E} = 1.5 and β_{lim} = 5% are selected according to available experimental and theoretical data, “Assumptions and Governing Equations: Ignition Condition” section. Those values were experimentally proven in W7AS and LHD respectively. The achievement of both values simultaneously is predicted for the W7X stellarator, “Assumptions and Governing Equations: Ignition Condition” section.
Concerning divertors, and considering the hypothesis and calculations in “Power Load on Divertor Targets” section, 30 MW/m^{2} thermal power load on targets is obtained for V = 30 m^{3}, Fig. 4. This power load is the practical limit for solid divertor targets [81, 82, 86], and a prospect for advanced Libased systems as divertor targets, “Lithium DivertorWall” section.
A Zamak alloy (a commercial alloy of zinc, aluminium, copper and magnesium) is selected for the coil support structures. Zamak is nonferromagnetic, easy to cast at low temperature (400–420 °C) in highprecision shapes, and has high yield strength S_{yield} = 360 MPa for the ‘Zamak 2’ alloy.
A strength safety factor of 1.5 accounts for uncertainties on the materials, stress concentration due to the ports and other uncertainties. From “Analytic Approximation of Stress” section and Eq. (12) with ψ = 0.5, it is calculated σ_{s} =240 MPa = S_{yieldZamak2} /1.5. However, σ _{max} (“Finite Element Calculation” section) exceeds S_{yield–Zamak2} . For Zamak 2 (E ≈ 85 GPa) the maximum displacement calculated by finite element analysis is 11 mm for ψ = 0.5. This displacement would be too large since coil positioning and shapes should have a tolerance of 0.1% or better [87, 88], corresponding to about 4 mm for iASTER. Therefore, it will be necessary to locally increase the thickness of the structure to ψ > 0.5 and to install a central support ring so as to balance the stresses and reduce the maximum displacement. These matters will be studied in future development stages.
i–ASTER.v1 specifications
Element  i–ASTER.v1  Ref. 

V  30 m^{3}  “Size and Materials for iASTER.v1 According to Limits” section 
B  9.8 T  “Assumptions and Governing Equations: Ignition Condition” section and Fig. 1 
R  3.8 m  
a  0.63 m  
A  6  “Assumptions and Governing Equations: Ignition Condition” section 
Plasma surface  95 m^{2}  
n _{line}  1.1 × 10^{21} m^{−3}  “Density and Temperature Needed for Ignition, Fusion Power” section and Fig. 2 
T _{0}  14.6 keV  
Fusion energy gain Q  Q → ∞ (ignition)  “Power Balance” section 
Fusion power  1.4 GW  “Density and Temperature Needed for Ignition, Fusion Power” and Fig. 3 
h_{E} (ISS04)  1.5  “Size and Materials for iASTER.v1 According to Limits" section 
<β >  5%  “Size and Materials for iASTER.v1 According to Limits" section 
τ_{E}  0.4 s  
Pulse length  2 s  5 τ_{E} 
Load on divertor targets (50% improvement, factor 2 sweeping, 50% radiation)  30 MW/m^{2}  “Power Load on Divertor Targets” section and Fig. 4 
Average neutron wall load  12 MW/m^{2}  “Power Load on Divertor Targets” section and Fig. 4 
Relative magnet thickness ε  1  “Power Dissipated in Resistive Magnets” section 
Weight of the copper magnet  ~ 1000 Ton  
Current per coil (one turn/coil, 144 coils)  1.6 MA  
Power consumed in the resistive copper coils  ~ 750 MW  “Power Dissipated in Resistive Magnets” section and Fig. 6 
Total magnetic energy stored  ~ 4.6 GJ  
Material of the monolithic support (initial selection)  Zamak 2  “Size and Materials for iASTER.v1 According to Limits” section 
Relative thickness of monolithic coil support Ψ  0.5  “Size and Materials for iASTER.v1 According to Limits” section 
Ave. stress on coil support at S  240 MPa  “Analytic Approximation of Stress” section and Fig. 9 
Max. local stress on coil support (QIP3 configuration, uniform Ψ)  600 MPa  
ΔT_{max} copper coils ~ insulation, only Ohmic (QIP3 ~ f_{c} = 5)  100 K  “Current Density and Coil Temperature” section 
Δ’ (distance LCFScoil)  0.3 m  “Distance from Plasma to Coils” section 
iASTER Specifications
Table 2 summarises the specifications of i–ASTER.v1.
Discussion of the Specifications
Lineaveraged plasma density up to n _{line} = 4 × 10^{20} m^{−3} was achieved in the High Density Hmode in W7AS [11] and a central plasma density of 10^{21} m^{−3} was reached in LHD [89]. The feasibility of n _{line} ~ 10^{21} m^{−3} should be experimentally proved, but, certainly, a highfield stellarator would favour high densities, according to the Sudo limit [21].
iASTER considers reactorrelevant β (5%) and adopts a high magnetic field B. As a result, the power density (∝ ~ β ^{2} B^{4}) and the heat load on the divertor is high. This will be an opportunity to test and enhance high power extraction systems and plasma purity, for example, by lithiumbased systems.
The evaluation of intermediate Q regimes and implications on the results (e.g. different divertor load) is beyond the scope of the present paper. These intermediate Q regimes might occur if ignition or nearignition could not be achieved in iASTER.
The electric power required for the magnets is substantial, but appears tractable. For example, TFTR flywheels provided up to 0.7 GW [90].
The use of steel would reduce the thickness of the monolithic structure. Nonetheless, steel requires more expensive casting and machining than Zamak. Alternatively, laminated composite (S_{yield} > 1000 MPa) shaped on additive manufacturing structures is envisaged, inspired by Refs. [69, 70].
Discussion on Neutronics
Neutron damage lower than 0.1 dpa is roughly estimated for the most exposed copper of the coils after 10 years lifetime (total of 1000 pulses, no shielding). This would produce some Cu embrittlement, but minor resistivity reduction and feasible insulation materials [91]. The estimation is based on the ratio r_{dpaNLW} of dpa per fullpoweryear (fpy) to the average neutron wall load (NWL), which is calculated from data in Refs. [92, 93] for ferritic–martensitic steels, resulting r_{dpaNLW} ~ 10 (dpa/fpy) / (MW/m^{2}). For the iASTER wall surface and total neutron power, with duty cycle 6 x 10^{−6}, ten years operation, peak NWL twice the average NWL [92], and dpa’s in copper 60% higher than in ferriticmartensitic steel [94], it results 0.03 dpa.
Concerning the neutron heating (‘nheat’) of coils, a first approximation is obtained as: i) the DEMO nheat at the first wall for ferriticmartensitic steel is taken, 8 W/cm^{3} [93], ii) nheat for copper and iron are similar [95], iii) scaling nheat to the plasma surface and neutron power in iASTER, with neutron shielding of 80%, resulting in nheat ~ 14 W/cm^{3}. For copper coil, an average ΔT_{aveNWL} ~ 8 °C is calculated at the end of the 2 s pulse (ΔT_{peakNWL} ~ 16 °C).
Regarding the nheat in the firstwall, following the previous procedure, without shielding, it results ΔT_{ave} ~ 40 °C (ΔT_{peak} ~ 80 °C).
No major neutronics difficulties are envisioned, thanks in part to the favourable high ratio of plasma surface to plasma volume in the relatively large aspectratio and small size i–ASTER.
Limitations and Discussion
Limitations
Different quasiisodynamic magnetic configurations (QIP3, HSR3) were utilized for the models. A definitive magnetic configuration for i–ASTER is not yet decided and it will have some impact on the resulting parameters. For example, the magnetic configuration impacts the areas of stress concentration (“Limitations and Discussion” section) and the current density factor (“Current Density and Coil Temperature” section).
Calculations by complex systems codes [25] have not been carried out yet, and will be the subject of future work. However, the rough estimates presented may be sufficient for this initial stage of development.
It is unknown if the assumptions performed for the estimation of the power load on divertor targets (large wetted area, sweeping, 50% edge radiation) can be simultaneously achieved. Lowering β to ~ 2.5% or less and increasing B could still achieve ignition at lower divertor loads.
The initial stress calculation does not tackle the insulation stress. Also, the (small) ports have not been modelled. The strength safety factor considered in the study may cover the uncertainties. However, further calculations will be required as the geometrical design advances.
Refined neutronics calculations are required to estimate the neutron damage to coil insulation, activation and damage on copper, and neutron heating of first wall and coils.
Discussion
A quasiisodynamic configuration was assumed for i–ASTER in order to advance the design. Currently, there is no universally accepted criterion to decide a best type of quasisymmetry, and it advises against an early decision on the definitive i–ASTER magnetic configuration.
Optimization of stellarator magnetic configurations continues worldwide [25, 96, 97, 98] and new stellarator concepts continue to emerge [96, 99]. Hence, future versions of iASTER might have larger A, which usually gives higher beta limit β_{lim} (“Assumptions and Governing Equations: Ignition Condition” section, [30]), or higher number of field periods. There is not any property (number of periods, type of quasisymmetry) of the selected QIP3 and HSR3 configurations that makes them unique for the mission and engineering approach of i–ASTER. Only, the intended small size of the device favours moderate aspect ratio.
Power extraction systems (e.g. solid divertor targets, flowing liquid metals) are critical for the attractiveness of fusion as a competitive energy source [47]. The liquid–metal option has been favoured for iASTER due to its high theoretical potential, e.g. high speed metal droplet beams [64] or molten tin shower jets [65], despite the comparatively limited level of development.
The massive resistive coils of variable crosssection involve new calculation methodologies that have only been initiated and represent a novel field of study.
Resistive magnets may not be the best option for stellarator power plants. Nevertheless, the requirement of simplification suggests this option for a first ignition experimental device.
If it is reasonable to study highfield ignitioncapable tokamaks like IGNITOR and FIRE, it appears reasonable to explore the potential of highfield stellarators of comparable size and magnetic field.
Summary and Conclusions
Wide ranges of physics and engineering parameters have been explored, in search for the conditions enabling ignition in a smallsize, highfield stellarator experiment. The magnets are resistive to contain construction costs. Specifically, massive copper coils of variable crosssection are envisaged to reach high fields with feasible power supplies. A monolithic toroidal coil support structure, external to the coils, is also proposed. Analytic expressions and finiteelement calculations were produced for the power consumed in the magnets and the stress in the monolithic support. Plots were generated for all the relevant parameters, under a variety of assumptions on the energy confinement enhancement factor h_{E}, stability beta limit β_{lim} and plasma volume. From this parametric study, a preliminary conceptual design of a highfield ignitioncapable experimental stellarator (i–ASTER) has emerged, based on a quasiisodynamic magnetic configuration. i–ASTER presents three distinctive features: massive resistive coils of variable crosssection, detachable periods and lithiumcoated walls and divertors. i–ASTER.v1 has a plasma volume of 30 m^{3} and an average magnetic field B ~ 10 T on axis, comparable with the IGNITOR and FIRE tokamak designs.
No unsurmountable difficulties have been found for this highfield pulsed stellarator approach to ignition experiments. The main concern is the possibly intractable power load on divertor targets and subsequent impurity influx. This could be tackled by lowering the operating β and using lithiumbased power extraction systems. The considerable radial thickness of the magnets is also a concern, but additive manufacturing could lessen this issue.
This work is undertaken in order to fill a gap in the knowledge of highfield ignitioncapable fusion devices of the stellarator type, which were significantly studied for tokamaks in the IGNITOR and FIRE tokamak concepts, and proposes a highfield resistivemagnet stellarator path towards the study of burning plasmas.
The definition and detailed calculation of the magnetic configuration and the 3D coil structure will be the subject of future work. Additive manufacturing of the coil support structure will also be further investigated. Detailed neutronics and more detailed mechanical and electric calculations will be performed in the next development stages.
Notes
Acknowledgements
The authors are grateful to M.I. Mikhailov, J. Nührenberg et al. [30] for supplying the QIP3 magnetic configuration, to A. Werner, J. Baldzuhn and J. Geiger for providing the coil definition of HSR3, and to E. Blanco and K.J. McCarthy for proof reading. The first author acknowledges J.A. Romero and J.A. Ferreira for longstanding discussions about fusion and stellarators. The work is partially funded by the Spanish ‘Ministry of Economy and Competitiveness’ under the grant number ENE201564981R (MINECO / FEDER, EU). This work is partly supported by the US Department of Energy under Contract DEAC0500OR22725 with UTBattelle, LLC and the US DOE.
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