Design, Construction and Initial Experiments of Internal-Coil Divertor Experimental Device SOLEIL

Abstract

A new torus device named SOLEIL, equipped with a copper internal coil (IC), has been constructed and plasma experiments have been initiated. A unique feature of SOLEIL is that the magnetic field structure of the divertor tokamak plasma, with either a single null point or double null points, can be generated without plasma current. Various shapes of the last-closed flux surface (LCFS) can be formed including both positive and negative triangularity. The primary objective of the device is to study the dependence of the scrape-off layer (SOL) flow and impurity transport in the SOL and divertor regions on the LCFS shape. In the design of the device, it was intended to make the IC as compact as possible, to minimize its impact on the plasma confinement. Dismountable toroidal field coils were also designed to be compact taking into account space constraints. The main parameters are as follows: the major radius of the IC is 0.18 m, the minor radius of the IC cover is 0.031 m, the nominal IC current is 8 kA⋅turns, and the nominal toroidal field is 0.0875 T at R = 0.18 m. The assembly was completed in January 2025 and the first plasma was generated in February 2025.

The accumulation of impurities has recently been recognized as a serious issue in magnetically confined plasmas. This is because tungsten is expected to be used as the plasma-facing material and impurities such as argon are expected to be injected into the divertor region to mitigate the divertor heat load in future fusion reactors. Impurity transport in the scrape-off layer (SOL) is as important as that in the core plasma. In the SOL, impurity transport is governed by forces such as friction with plasma flow and thermal forces [1]. Various divertor configurations, including those with negative triangularity, have been proposed, and the dependence of SOL flow and impurity transport on plasma shape needs to be evaluated. In some models of SOL plasma flow, ions crossing the last closed flux surface (LCFS) due to toroidal drift play an important role [2, 3], and this process is influenced by the plasma shape.

A small device with a variety of divertor configurations is useful for studying plasma flow and impurity transport in the SOL because of its excellent diagnostic accessibility. However, it is difficult to sustain vertically elongated plasmas with a poloidal divertor in small devices due to the need for feedback control of the plasma vertical position. Another issue is the short duration of plasma current in small devices due to the limited available flux. To study SOL plasma properties in divertor configurations without these issues, an experimental device called SOLEIL (Scrape-Off Layer Experiment with Internal coiL) was planned and constructed at Nagoya University. In this device, tokamak-like magnetic field configurations are generated using the current in an internal ring copper coil instead of plasma current, in addition to external equilibrium field coils and the toroidal field (TF) coils.

A schematic cross-sectional view is shown in Fig. 1. This device is installed in place of the small tokamak device TOKASTAR-2 [4], and the cylindrical part and the top plate of its vacuum vessel are reused. Both the radius and the height of the cylindrical part are 600 mm. A center stack was newly manufactured. It consists of a center post, 56 copper conductors, and other components. The center post has a cylindrical shape with a diameter of 114 mm and is equipped with covers that protect the feeders. The conductors are used as the inner legs of the TF coils. The major radius (R) of the internal coil (IC) was set to 180 mm, approximately midway between the inner wall at R = 72 mm and the outer wall at R = 300 mm.

Fig. 1.  Cross-sectional view of the conceptual configuration of the SOLEIL device. Red lines indicate magnetic flux surfaces. The green square and six brown squares denote the internal coil and the equilibrium field coils, respectively. The outermost yellow rectangle represents the TF coil.

A key requirement in the design of the TF coils was that they should be demountable, i.e., mechanically assembled and disassembled, while minimizing their height. Although wide plates are used in the spherical tokamak LATE for the return section [5, 6], only 10–20 cm of vertical space was available in SOLEIL due to the room height constraints.

The conductors in the center stack have a cylindrical shape with a diameter of 10 mm and are fixed by resin impregnation. The structure of the center stack is similar to that of LATE, although the conductors are not water-cooled in SOLEIL. A thermocouple is attached to one of the conductors near the equatorial plane to monitor the conductor temperature during experiments and during vacuum vessel baking.

The return section of the TF coils has eight legs, each with seven turns. Each leg consists of an upper radial part, a vertical part, and a lower radial part. These parts and the central conductors are connected with M5 screws, as shown in Fig. 2. Each leg has three layers with 3, 2 and 2 turns from the inside. The conductors have a square cross-section of 10 × 10 mm. Their surfaces are almost entirely covered with glass tape for electric insulation because the gap between adjacent conductors is only 2 mm. The rated TF coil current is 1.41 kA (78.8 kA⋅turns), which generates a magnetic field of 0.0875 T for electron cyclotron resonance (ECR) of 2.45 GHz microwave at the position of the IC, R = 180 mm.

Fig. 2.  Connection of TF coil conductors (a) between the central conductors and the upper part and (b) between the upper part and the vertical part.

A key requirement in the design of the IC and its supports was to minimize the interaction with the plasma. The design was based on the Proto RT device [7]. The IC consists of 285 turns of 2 mm diameter conductors and can carry approximately 8 kA⋅turns. The cross section of the IC with the cover is shown in Fig. 3. The winding has a square shape of 35 × 34 mm with chamfered corners. This shape minimizes the size of the IC cover. Four support components are located on the low-field side to efficiently use the space between the winding and the cover, since the center of the flux surface is shifted outward relative to the conductor center in typical configurations. Two stainless steel rods (1.5 mm diameter) are attached to each support component. The IC is supported by a total of eight support rods (four upper and four lower) connected to the center post. Each rod is covered with an alumina pipe (3 mm diameter). The diameters of the rod and the pipe were minimized while maintaining mechanical strength. The IC is protected by a torus-shaped cover with a cross-sectional diameter of 62 mm, achieved through design optimization. The cover consists of eight parts with unsealed gaps. It is electrically insulated from the coil and supports to allow the application of a radial electric field. The IC feeder and the voltage supply wire for the cover are routed from the high-field side of the IC, between the upper and lower cover sections and are guided into the center post cover. They are covered with an alumina pipe of 7 mm diameter.

Fig. 3.  Cross-sectional CAD view of the internal coil: the winding (pink), support component (sky blue), support rods (brown), and the coil cover (yellow).

Three pairs of equilibrium field (EF) coils are installed on the upper and lower sides of the device: two pairs inside the vacuum vessel and one pair outside (see Fig. 1). The outer pair is reused from TOKASTAR-2. Upper and lower divertor plates made of stainless steel are installed at ± 180 mm from the equatorial plane. They are electrically insulated from the vacuum vessel and from each other, and voltage can be applied via feedthroughs.

Magnetic flux surfaces were calculated by tracing field lines based on the Biot-Savart law. Various divertor configurations with different triangularities and numbers of null points can be obtained by adjusting the EF coil currents. An example of lower single-null configuration with nearly zero triangularity is shown in Fig. 1. The plasma triangularity $\delta$ can also be varied over a wide range, including both positive and negative values as shown in Fig. 4. The shape parameters are: ${\delta }_{95}$ = 0.25, ${\delta }_{\mathrm{X}}$ = 0.35, ${\kappa }_{\mathrm{X}}$ = 1.61, $A$ = 2.57 in (a), ${\delta }_{95}$ = 0.01, ${\delta }_{\mathrm{X}}$ = 0.00, ${\kappa }_{\mathrm{X}}$ = 1.81, $A$ = 2.45 in (b), and ${\delta }_{95}$ = −0.18, ${\delta }_{\mathrm{X}}$ = −0.36, ${\kappa }_{\mathrm{X}}$ = 1.64, $A$ = 2.63 in (c), where ${\delta }_{95}$ and ${\delta }_{\mathrm{X}}$ denote the triangularity at the 95% poloidal flux surface and at the separatrix, respectively, and ${\kappa }_{\mathrm{X}}$ and $A$ denote the elongation and aspect ratio at the separatrix. The safety factor ($q$) profile for Fig. 4(b) is shown in Fig. 5 for IC current 8 kA⋅turns and toroidal field 0.0875 T at R = 0.18 m. Since the toroidal current is concentrated in the IC, the poloidal field is approximately inversely proportional to the minor radius, resulting in strong positive magnetic shear and low $q$ near the IC. This strong poloidal field is expected to reduce the plasma loss to the IC cover.

Fig. 4.  Examples of double-null configurations with (a) positive triangularity, (b) near-zero triangularity, and (c) negative triangularity.

Fig. 5.  Example of the safety factor ($q$) profile. The horizontal axis represents the horizontal minor radius.

Plasma generation and heating are performed using 2.45 GHz microwaves injected through an equatorial port. The planned plasma duration is approximately 100 ms since all coils are inertially cooled. The adiabatic temperature rise rates are estimated to be 1.7 K s⁻¹ for the TF inner legs and 0.4 K s⁻¹ for the IC at nominal currents. All coils except the TF coils are powered by DC supplies. The working gases, hydrogen, helium, nitrogen, or argon, are continuously fed into the vessel through a port behind the top divertor plate.

Manufacturing of parts and components including TF coil conductors began in February 2023. Assembly began in May 2024. TOKASTAR-2 was disassembled in October 2024. Assembly of SOLEIL was completed and the vacuum pumping was started in January 2025. Figure 6 shows the device during assembly. One can see the center stack, the internal coil partially covered, half of the lower divertor plates, two lower EF coils, the bottom plate of the vacuum vessel and the lower part of the TF coils.

Fig. 6.  Photograph of partially assembled in-vessel structure of the SOLEIL device.

In initial experiments, the TF coils were driven by capacitor banks (5 mF, 1 kV) with a crowbar diode and a thyristor switch. The current was measured with a shunt resistor (50 μΩ). The peak current exceeded the nominal current 1.41 kA with the charging voltage of 960 V. The waveforms of the current and the voltage of the TF coil are shown in Fig. 7. The coil inductance (1.5 mH) and the coil resistance (55 mΩ) were evaluated from these waveforms. The inductance agrees with the design value while the resistance is about 1.7 times higher, likely due to contact resistance.

Fig. 7.  Waveforms of the voltage (red) and the current (blue) of the TF coils.

In February 2025, the first plasma was generated using only TF coils and injecting microwaves for 5 ms with argon gas. Subsequently, experiments using IC and EF coils were conducted. Figure 8 shows the device and the plasma, where one can see the IC in the plasma. In initial experiments, upper and lower coils of each of in-vessel EF coil pairs were connected in series due to limited power supplies, resulting in double-null configurations.

Fig. 8.  Photograph of the SOLEIL device with plasma.

Gas pressure scans (Fig. 9) showed that plasma generation required a pressure larger than 0.04 Pa with toroidal field only, whereas with IC and EF coils, plasma was generated at 5 mPa and even at 2.7 mPa without gas feed. This demonstrates improved electron confinement with closed flux surfaces.

Fig. 9.  Plasma generation rate as a function of gas pressure. The internal coil current is 7.1 kA⋅turns.

An example of waveforms in the plasma experiment is shown in Fig. 10, together with the flux surface shapes. The working gas was helium. A 1.5 kW microwave was injected from t = 3 to 18 ms, during which the ECR layer moved between R = 0.20 and 0.14 m. The plasma light intensity was nearly constant during that period. A fast camera image of the same discharge at t = 15 ms is shown in Fig. 11. A ring-shaped plasma surrounding the IC is observed.

Fig. 10.  (a) Time evolution of the major radius of ECR layer, injected microwave power and plasma light intensity measured with an avalanche photodiode. (b) Corresponding flux surface shapes. The internal coil current is 6.8 kA⋅turns.

Fig. 11.  Fast-camera image of plasma with closed flux surfaces.

In summary, the SOLEIL device with a compact internal coil and dismountable compact toroidal field coils has been successfully constructed, and the plasma experiments have begun.

The authors thank Dr. H. Tanaka for giving them information on the TF coils of LATE and thank Dr. H. Saitoh and Mr. J. Morikawa for giving them information on the internal coil of Proto RT. This research was supported by JSPS KAKENHI 22H01202, 23K22473 and the NIFS Collaboration Research program No. NIFS23KIPP029.

References

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