Magnetic Field and Force of Helical Coils for Force Free Helical Reactor (FFHR)

The electromagnetic force on a helical coil becomes smaller by decreasing the coil pitch parameter which is the angle of the coil to the toroidal direction. This makes it possible to enlarge the central toroidal field or to simplify the supporting structures of the coil. The plasma minor radius, however, becomes smaller with the pitch parameter, and a higher field is necessary to attain the same plasma performance. Another important item in a helical reactor is the distance between the helical coil and the plasma to gain enough space for blankets. In order to reduce the mass of the coil supports, a lower aspect ratio is advantageous, and an optimum value of the pitch parameter will exist around 1.2 and 1.0 for the helical systems of the pole numbers of 2 and 3, respectively

Onset of instability with collapse observed in relatively high density and medium beta regions of LHD

Edge MHD instabilities with pressure collapse are found in relatively high beta and low magnetic Reynolds number regions with a magnetic axis torus outward-shifted configuration of the large helical device (LHD), and characteristics and onset conditions of the instability are investigated. The instability has a radial structure with an odd parity around the resonant surface, which is different from that of the interchange instability typically observed in the LHD. The onset condition dependence on the magnetic axis location shows that the onset beta increases as the magnetic axis location moves more torus inwardly, and the instability appears only in limited configurations where the magnetic axis is located between 3.65 and 3.775 m. In such configurations, the resonant surface location is close to an index of the plasma boundary. This fact suggests that the distance between the resonant surface location and the plasma boundary plays an important role in the onset, and a possibility that the instability is driven by an external mode

Developments of frequency comb microwave reflectometer for the interchange mode observations in LHD plasma

We have upgraded the multi-channel microwave reflectometer system which uses a frequency comb as a source and measure the distribution of the density fluctuation caused by magneto-hydro dynamics instability. The previous multi-channel system was composed of the Ka-band, and the U-band system has been developed. Currently, the U-band system has eight frequency channels, which are 43.0, 45.0, 47.0, 49.0, 51.0, 53.0, 55.0, and 57.0 GHz, in U-band. Before the installation to the Large Helical Device (LHD), several tests for understanding the system characteristics, which are the phase responsibility, the linearity of output signal, and others, have been carried out. The in situ calibration in LHD has been done for the cross reference. In the neutral beam injected plasma experiments, we can observe the density fluctuation of the interchange mode and obtain the radial distribution of fluctuation amplitude

Experimental study of non-inductive current in Heliotron J

It is important to control non-inductive current for generation and steady-state operation of highperformance plasmas in toroidal fusion devices. Helical devices allow dynamic control of non-inductivecurrent through a wide variety of magnetic configurations. The reversal of non-inductive current consisting of bootstrap current and electron cyclotron driven current in electron cyclotron heating plasmas has been observed in a specific configuration at low density in Heliotron J device. By analyzing thenon-inductive current for normal and reversed magnetic fields, we present experimental evidence for the reversal of bootstrap current. Our experiments and calculations suggest that the reversal is caused bya positive radial electric field of about 10 kV/m. Moreover, we show that the typical electron cyclotron current drive efficiency in Heliotron J plasma is about 1.0 × 1017 AW?1m?2, which is comparable to other helical devices. We have found that the value is about 10 times lower than that of tokamak devices. This might be due to an enhanced Ohkawa effect by trapped particles

Calibrations of Fast Ion Flux Measurement Using a Hybrid Directional Probe

A hybrid directional probe method both “thermal and Langmuir probe” was applied for fast ion measure- ments in the compact helical system. In order to obtain absolute values of fast ion density and power density, a calibration of the probe was performed using neutral hydrogen beam and a mixture beam of hydrogen and proton, of which beam current and energy were controlled. The conversion factor from temperature increase of the probe head to local power density and secondary electron emission yield was obtained. The density of fast ions was obtained by directional thermal probe (DTP) method inside the last closed flux surface, and the density ratio was nFastIon/nBulkPlasma = 2.7 × 10?3 at r/a = 0.9. The observation of the directional Langmuir probe (DLP) method is consistent with the DTP results

Drift stabilization of ballooning modes in a high-⟨β⟩\langle \beta \rangle LHD configuration

Ideal MHD yields at best inconclusive predictions about the stability of the LHD heliotron for $\langle \beta \rangle \geq 3\%$. We investigate the impact of the drift stabilization of ballooning modes for the inward shifted LHD configuration (vacuum magnetic axis $R_0 \sim 3.5m$). The background equilibrium is considered anisotropic in which the neutral beam ions contribute about $1/4$ fraction of the total diamagnetic beta, $\langle \beta_{dia} \rangle$. A drift corrected ballooning mode equation obtained from the linearized gyrokinetic equation is expanded assuming that the hot particle drifts are much larger than the mode frequency. The fast particle pressure gradients contribute weakly to both the instability drive and the diamagnetic drift stabilization (which is dominated by the thermal ion diamagnetic drifts) for $\langle \beta_{dia} \in [0,4.8] \%$. In the single fluid limit (diamagnetic drifts ignored), the thermal pressure gradients drive ballooning modes in a broad region encompassing the outer $60-90 \%$ of the plasma volume at $\langle \beta_{dia} \rangle \approx 4.8 \%$. To stabilize these modes, we find that diamagnetic drift corrections must be invoked (mainly due to the thermal ions). The energetic ion diamagnetic drifts play a role only for low wave number values, $k_{\alpha}\leq 8$. It has been verified that the fast particle drift ordering imposed by the model is amply satisfied for on-axis hot particle to thermal density $N_{h0}/N_{i0} \approx 1\%$ even at high $\langle \beta_{dia} \rangle$

External RMP effect on locked-mode-like instability in helical plasmas

The slowing-down mechanism of the locked-mode-like instabilities with and without an island structure is investigated through the effects of an external RMP (resonant magnetic perturbation) on the instabilities. For both instabilities, the slowing-down duration decreases with the increase in the external RMP, and the RMP dependence is consistent with the braking model of the j × B force due to the interaction between the instabilities and the external RMP. Moreover, the relationship between the amplitude and the frequency of both locked-mode-like instabilities during the slowing down is consistent with the force balance model between the j × B force due to the external RMP and a viscous force. These results suggest that the slowing down of both locked-mode-like instabilities with finite external RMP occurs due to the j × B force driven by the external RMP

Development of a Hierarchy-Integrated Simulation Code for Toroidal Helical Plasmas, TASK3D

The present status of the development of a hierarchy-integrated simulation code for toroidal helical plasmas, TASK3D, is reported. TASK3D is developed by extending the integrated modeling code for tokamak plasmas, Transport Analyzing System for tokamaK (TASK) [A. Fukuyama et al., Proc. of 20th IAEA Fusion Energy Conf. (Villamoura, Portugal, 2004) IAEA-CSP-25/CD/TH/P2-3]. In order to extend TASK to be applicable for threedimensional configurations, a new module for the radial electric field in general toroidal configurations has been developed and implemented. As a first test for this implementation, numerical simulations for the time evolution of temperature and electric field are conducted on the basis of an LHD experimental result, by a successful combination of a diffusive transport module and the implemented electric field module