Development of high-power heating for large helical device

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Fusion Engineering and Design 26 (1995) 227-238

Fusion Eng! ng aridDesign

Development of high-power heating for large helical device T. Kuroda, K. Ohkubo, T. Watari, M. Sato, R. Kumazawa, T. Mutoh, O. Kaneko, Y. Oka, S. Kubo, Y. Takeiri, K. Tsumori, A. Ando, H. Idei, T. Seki National Institute for Fusion Science, Nagoya 464-01, Japan

Abstract

Three high-power heating mechanisms, i.e. electron cyclotron resonance heating, ion cyclotron resonance heating and neutral beam injection heating will be available to achieve operation of the Large Helical Device in the (i) high nrT mode, (ii) high Ti mode and (iii) high fl mode. The requirements for each heating method are (i) 10 MW of ECH at a frequency of 84 GHz with a pulse length of 10 s, (ii) 3 MW of ICRF heating in the frequency range 25-100 MHz in CW operation and (iii) 20 MW of NBI at a hydrogen injection energy of 125 keV and 10 s operation. The components of each heating mechanism which need development individually exist, i.e. (i) a high-power gyrotron with an output power of 1 MW in CW operation and a high-efficiency transmission line for ECH, (ii) a high-power co-axial transmission line with cooling and an automatic matching circuit for ICRF heating and (iii) a large negative ion source and some components of the beam line for NBI. The heating scenario and the present status of the development are described for each heating system.

I. Introduction

A helical device, in which the confinement magnetic field is produced by external windings, is attractive as an alternative concept to the tokam a k device. In order to clarify the important physics and engineering issues in designing a future helical reactor (by studying the behaviour of the currentless plasmas in large-scale experimental helical devices), a superconducting large helical device ( L H D ) [1] is under construction at the National Institute for Fusion Science (NIFS). The purposes of the L H D project are as follows: (a) to investigate transport in high n z T plasmas ( T i = 3 - 4 k e V , n e = 1 0 Z ° m -3, Ze=0.1--

0.3 ms, B = 4 - 3 T) extrapolatable to reactor plasmas; (b) to achieve a high-beta plasma ( ( / ~ ) > 5 % , B = 1 - 2 T) as needed for the reactor, and to understand the related physics; (c) to obtain basic data necessary to realize steadystate operation through experiments on the quasi-steady plasma control using divertors; (d) to study the behaviour of high-energy particles in the helical magnetic field and to conduct particle simulation; and (e) to increase the comprehensive understanding of toroidal plasmas by carrying out studies complementary to those in t o k a m a k devices. In addition to these physical tasks, the other important subjects are technological innovations

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T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227 238

such as the development of high power steadystate heating mechanisms. To achieve these plasma parameters, we are planning to use 10 MW of ECH at a frequency of 84 GHz in CW operation, 3 MW of ICRF heating in the frequency range 25 100 MHz in CW operation and 20 MW of NBI at a hydrogen injection energy of 125 keV in 10 s operation. In this paper, the heating scenario of the LHD and the present status of the development of these heating systems are overviewed.

2. Heating scenario of LHD In the LHD, the initial target plasma will be produced by ECH and then will be heated by intense high-power additional NBI and ICRF heating. The main requirements for the heating are (1) to produce a good plasma at a variety of magnetic field strengths, (2) to obtain the predicted plasma parameters over the wide range of operating conditions, (3) to realize long pulse or steady-state heating, (4) to control the plasma electric potential for the improvement of the confinement and (5) to produce an energetic ion tail for the study of alpha particle simulation. 2.1. Electron cyclotron heating

The main purpose of ECH is to produce a target plasma reproducibly and efficiently. A frequency of 84 GHz corresponds to a resonant magnetic field of 3 T. The cut-off density of the fundamental ordinary mode (O-mode) is about 8.8 × 1019 m -3. Since a right-hand cut-off layer and a resonance layer of upper hybrid exist in the region between the launcher and the centre of plasma, only the ordinary waves (O-waves) can penetrate. Then it is necessary not only to control the electric field of the wave but also to rotate the polarization surface in such an experiment on improved confinement due to control of the magnetic surface position and the vertical magnetic field. Controllability of the power deposition profile is a useful tool for the investigation of plasma transport by controlling the injection angle of the wave. On the other hand, the cut-off

density of the second harmonic X-mode is about 1.7 x 1019cm -3. The second harmonic heating scheme becomes feasible in addition to the fundamental heating mode. If the ECH system has two frequencies, it will give a wide range of operational plasma parameters in experiments in the LHD. 2.2. I C R F heating

ICRF heating is considered as a steady-state additional heating mechanism, to achieve the objective plasma parameters and as a useful tool for studying many aspects of physics in the LHD. There are several candidate wave modes which can heat the plasma effectively. Methods of twoion hybrid heating, minority heating, higher harmonic fequency heating and ion Bernstein wave heating have features corresponding to electron heating, ion heating, high-energy tail particle production and bulk ion heating, respectively. Ion Bernstein wave heating has the merit that there is no high-energy tail production. Fast-wave heating is the most established method to heat plasmas and to produce a high-energy ion tail which will provide an alpha particle simulation tool on the LHD. Two-ion hybrid heating and minority heating as the main ICRH methods are also candidates for heating LHD plasmas. By using three antenna systems for fast-wave excitation, it is planned that an ICRF power of 3 MW in CW operation or 9 MW in 10 s operation will be applied to heat the plasma. To maintain the high efficiency and long pulse operation, it is necessary to suppress the orbit losses of the energetic ions and parasitic loading due to the SOL plasma near the antennas. The present concept of the antenna design is the most plausible for solving these problems of ICRF heating in the helical device. The antenna is placed on the r - O plane and the outward side of the torus, where the plasma cross-section is vertically elongated as shown in Fig. 1. The cyclotron resonance layer is settled at the inward side of the torus. In this design, the accelerated ions are well confined and the loss region in the velocity space is minimized. At the same time, there is enough space to set the large antenna loops, protectors, electrostatic

T. Kuroda et al. [ Fusion Engineering and Design 26 (1995) 227-238

Fig. 1. Schematicdiagram of high-fieldand outward side ICRF antenna. screens and cooling structures. In addition, there is no cyclotron resonance layer near the antenna, and the local magnetic shear is relatively small. These conditions are beneficial for obtaining a high loading resistance such as 5 - 1 0 f~.

229

is about 30%. The former is increased gradually and the latter is increased up to 45% with increasing beam energy. We concluded from these results that tangential injection is needed to achieve efficient heating and core plasma heating of above 50% in both the low and high plasma density modes achieved at an injection energy of 100150 keV. Taking the shine-through power in the low-density mode into account, the injection energy is fixed between 100 and 130 keV. On the other hand, the injection port of the L H D is relatively small, i.e. about 40 x 40 cm 2. Therefore, to reduce geometrical loss of beams to less than 25%, the beam divergence angle is required to be less than 0.5 ° . To satisfy both conditions with regard to injection energy and beam divergence angle, the negative-ion-based NBI has been adopted. A total power of 20 M W at a hydrogen beam energy of 125 keV is injected with four tangential injectors.

3. Research and development of each heating system 3.1. R & D for ECH system

2.3. Neutral beam injection heating The optimization of neutral beam injection heating has been investigated with respect to the ratio of the thermalized power within a half radius to the port-through power as a function of the injected angle, and the ratios of the absorption power within a half radius to the port-through power as a function of the injection energy (H ° beam) at the tangential injection [2]. In the prependicular injection, the orbit loss of the total beam attains 80% and the thermalized power is few percent, although a beam of 40% is deposited in the plasma. In the tangential injection the orbit loss is less than 10% for co-injection and the thermalized power is nearly 60%, while an injected beam of 60% is deposited within the half radius. The absorbed power of the beam is nearly 50% at the energy above 100 keV and the shine-through power is less than 5% with a high nr Ti plasma. On the other hand, in the high Ti mode the absorbed power is nearly 50% and the shine-through power

Research and development on the ECH system for the L H D have been carried out for an MWclass gyrotron and a high-efficiency, high-power r.f. transmission line, which are the most important engineering issues for realizing the ECH system for the L H D device. The first step of R & D for the gyrotron started with a 84 G H z whispering gallery gyrotron, the specifications of which are shown for a prototype and for a developing system in Table 1. The gyrotron is designed to generate an output power of 1 M W in 10 s and 0.8 MW in CW operation. The design of the gyrotron is shown in Fig. 2. The tube is designed to extract the spent electron beam from the gap between wave guide on the output. The tube demonstrated successful operation of a long pulse at 500 kW output power and a short pulse at 900 kW, as shown in Fig. 3. The output efficiency could be kept nearly constant by optimizing the magnetic field configuration. The specific heat load on the collector is less than

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T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227-238

Table 1 Specification of the 84 GHz whispering gallery gyrotron Type Gun Frequency Output mode

Whispering gallery mode Magnetron injection gun 84 + 0.5 GHz TE~5.2 which will be mode-transformed to TEMoo Mode purity > 95% Pulse width 1 MW, 10 s design value 0.8 MW CW design value Efficiency Minimum30% Type Microwave-beam separation Voltage 80 + 0.2 kV Beam current Maximum 62 A SC magnet 3-4 T

almost all the power can pass through the output window. A good gaussian beam pattern is observed at the vacuum window (Fig. 4). The output microwave beam intensity at the window will be taken into account so as to reduce the heat and mechanical problems in high-power CW operation. A high-voltage regulated power supply with 50 A in 10 s and 25 A in CW operation at 80 kV is constructed. The power supply has two deck modulator tubes, a series inductor to limit the rate of current rise in the gyrotron arcing and a crowbar circuit to protect the gyrotron. In the real E C H transmission system for the L H D , we are planning a suitable combination of the quasi-optical mirrors and the conventional circular corrugated waveguide system with H E of a hybrid mode. Considering the flexibility of the transmitting frequency mode, path and the ability to vary the injection angle and focusing point, we started the R & D p r o g r a m m e on the transmission system by studying the mirror type. In addition, the performance of a circular and rectangular corrugated waveguide component, bend and polarizer were experimentally checked by using the developed methods which measure the mode content and radiation pattern. Fig. 5 shows a preliminary design of the E C H antenna system for the L H D . In the normal poloidal cross-section of the L H D where the

II

0.5 k W cm -2 at a beam current of 22 A when the output power is 500 kW. The extrapolated collector heat load in the 1 M W C W condition is still safe and low enough to handle. Hence the results indicate that the designs of the electron gun, beam shaver, cavity and collector seem to be encouraging. In the second step of development, the gyrotron will be replaced by a new gyrotron with a built-in converter as shown in Fig. 4. The built-in converter consists of a modified Vlasov-type launcher with visor and beam-shaping mirror. Cold testing of the converter has been carried out, and modification of the converter has been repeated until

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T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227 238

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mode B surface (constant total magnetic field surface) is complicated as compared with that of the tokamak, a highly focused ECH beam is required to obtain a well defined power deposition profile and to obtain clear results. Further, the superconducting helical coils surrounded by the bell jar mean that the port access is far from the plasma and a special beam must be designed to obtain a narrow power deposition profile. The

allowed beam size is limited by the port space and long duct to the L H D vacuum vessel. An elliptical gaussian beam satisfies both requirements of the port space and sharp power deposition. As another method to obtain an elliptical gaussian beam, we developed a parallel corrugated plate system connected to the circular corrugated wave guide, as presented in another paper at this conference [3].

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T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227-238

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3.2. R & D f o r I C R F heating

Since I C R F heating is considered as the main additional heating in a steady-state plasma experiment, technical developments of I C R F heating were carried out to meet the requirement of highpower steady-state heating in the test assembly. The components to be developed included a steady-state power amplifier, water-cooled transmission lines, fast impedance matching circuits and a wave guide IBW launcher. Fig. 6 shows an outline of the r.f. system for the L H D . The system consists of four units each of which contains a 3 M W r.f. source, a tuner and two antenna elements.

In order to carry out the development study easily and efficiently, a test assembly [4] was constructed at the Toki site. Centred by a vacuum chamber in which various antennas are to be tested, it is furnished with a high-power amplifier, a d.c. break and a stub tuner. These components will become prototypes of those of the L H D I C R F after some improvements and final confirmation of successful performance. The amplifier has been tested up to 400 k W for 10 min and is waiting for further testing up to 1.5 MW. The components of the assembly have been tested at a power level of 30 kW. A test antenna (40 cm 2 × 70 cm) was installed in the vacuum chamber. The stub tuner was able to match the impedance. The d.c. break which is installed between the two stub tuners has been designed for an insulated voltage of 10 kV and a level of r.f. leakage power of less than 80 dB at 50 MHz. The results of testing the break show that it seems to be well designed since no problem of r.f. leakage arose. Because the L H D I C R F heating system will be operated in the steady state, all the components need to be water cooled. In order for assembly into a coaxial transmission line, the development of the junctions of the various water-cooled components is essential, because they have to be connected electrically tight and completely free from water leakage. So far, we have been operating the assembly at a power level of 30 kW and no problems were observed in the transmission line. This power level corresponds to a 500 kW power injection in a forthcoming L H D experiment with a plasma load. Particular attention was paid to the temperature rise of the antenna, which can be seen through the window of the vacuum chamber. Temperatures at various points on the antenna were measured using a far-infrared camera. As shown in Fig. 7, the carbon lateral protectors showed a considerable temperature rise. The extrapolation of these data may suggest that the present model antenna needs some improvement for heat removal from the lateral protectors. A waveguide antenna [5] has been studied for application to IBW heating. A Type-III antenna [6] was used in CHS to demonstrate the principle of Ion Bernstein wave heating. After its usefulness

7". Kuroda et al. / Fusion Engineering and Design 26 (1995) 227-238 transmission lines

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with ORNL. The original folded waveguide antenna has a blind in every two sections (polarization plates). In our modified F W G A for IBWH, the polarization plates are removed in order to give built-in k If phasing favourable for slow wave excitation A schematic drawing of F W G A for L H D is shown in Fig. 8.

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A conceptual design of a negative ion-based neutral beam injection in the L H D is shown in Fig. 9. The beam line has two ion sources. The design parameters of the 125 keV and 2 0 M W negative ion-based NBI system for L H D are given in Table 2. The most important engineering issue for the negative ion-based NBI is a large high-current negative ion source. The performance requirements for a large negative ion source are as follows: (1) high negative ion current density of more than 30 mA cm-2; (2) low operating pressure of less than 1 Pa; (3) low stripping loss at the extraction grid; and (4) low ratio of extracted electron current to negative ion current. The development of the negative ion source was started by the optimization of 1/6- and 1/3-scale negative ion sources [7,8], the specifications of

T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227 238

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T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227 238 Table 2 Design parameters of 125 keV and 20 M W negative ion-based NBI system for L H D Overall Injection power Beam energy Pulse length Ion species N u m b e r of injectors N u m b e r of ion sources Injection angle

Neutralizer 20 M W (through port) 125 keV (H) 250 keV (D) 10 s H/D 4(H)/2(D) 2 per injector Tangential injection (balanced injection)

Size

25 x 150 x 5 0 0 c m 3

Line density

16.7 Pa cm (4.4 x 10 ~5 molecules cm -2) 6.5 x 10 2 Pa (at entrance) 1.5 x 10 3 p a (at exit)

Pressure

Cryo-pump Source chamber Beam d u m p chamber Refrigerator power

505 m 3 s l 2040 m 3 s 1 k W (at 3.7 K)

Ion source Injection port Source type Grid size Transparency Extracted current Current density Beam divergence angle

Volume production 25 x 150 cm 2 ( 3750 cm 2) 40% 45 A 30 m A cm 2 0.5 °

Size Length Pressure

400 500 m m diameter 3m 1.5 x 10 3 Pa

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235

T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227-238

236

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Table 3 Specification of test stand Ion source

Acceleration energy Pulse duration A t o m species

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P-arc (kW) which are given in Table 4. A schematic diagram of the 1/3-scale negative ion source is shown in Fig. 10. A volume production bucket negative ion source is adopted. The size of the rectangular plasma source is length 62.5 cm, width 37 cm and depth 17 cm. The beam acceleration system consists

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Fig. 11. Extracted total hydrogen negative ion current as a function of the discharge power of the plasma source.

of four grids, i.e. plasma, extraction, electron suppression and ground grids. Each grid is divided into two subgrids and are inclined to each other in

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T. Kuroda et al. / Fusion Engineering and Design 26 (1995) 227 238

237

Table 4 Specification of 1/3 and 1/6 scaled ion sources Parameter

Full-scale

1/3-scale

l/6-scale

Beam energy

125 keV (H) 250 key (D) 45 A 10 s 25 x 150 cm2 (3750 crn2) 9 mm diameter ( 2360 holes) 40% 30 mA cm-2 0.5"

120 keV (H)

120 keV (H)

15A ls 25 × 50 cm2 ( 1250cm2) 9 mm diameter (800 holes) 40% 30 mA cm 2 0.5°

7.5A ls 25 × 25 c m 2 (625 cm2) 9 mm diameter (400 holes) 40% 30 mA cm -~ 0.5"

Beam current Pulse length Grid size Hole size Transparency Current density Beam divergence

order to focus the ion beam. The negative ions are extracted from 800 holes of 9 mm diameter and focused on the focal point 6 mm downstream from the ion source. The source is operated with caesium seeding at a pulse duration of 1 s. Fig. 11 shows the total hydrogen negative ion current as a function of the discharge power of the plasma source. A total hydrogen negative ion current of 10.5 A is obtained at an arc power of 170 kW but it is still less than the design value of 15 A. The extracted beam current is measured calorimetrically. Another important engineering issue for negative ion-based NBI is beam line components such as ion and electron beam dumps, high pumping speed cryopump and calorimeter. To develop these beam line components and a prototype negative ion source and to study efficient beam transport, a neutral beam test stand [9] has been constructed on the basis of conceptual design of the L H D NBI. A schematic drawing of the test stand is shown in Fig. 12 and the specification is given in Table 3. It consists of two large vacuum tanks and a long gas neutralizer. The ion source vacuum tank has 45 m 3 s-~ high pumping speed cryopump to reduce stripping losses at the extraction grid. A vacuum-immersed large negative ion source is being examined but it is still at a preliminary experimental stage.

4. Conclusion The high-power plasma heating on the L H D has been described and the development of the

hardware for three heating systems, i.e. ECH, I C R F heating and NBI, has been outlined. Concerning ECH, a high-power gyrotron with design values of an output of 1 M W in 10 s and 500 kW in CW operation was prepared and it is being tested for higher output power and long pulse operation. A second tube with a built-in converter has been designed and is now being fabricated. The transmission line of ECH, a combined system of quasi-optical mirror and waveguide, has been tested at low power levels. A steady-state operation test has been carried out on the water-cooled junction and components of I C R F system. Encouraging results were obtained for steady-state operation and the feasibility of a waveguide antenna has been demonstrated in a CHS experiment. A high-current hydrogen negative ion beam of 10.5 A was achieved in Cs seeding operation of the plasma source. A neutral beam test stand for development of the beam line components of an L H D injector has been constructed and a vacuum-immersed large negative ion source is being examined. The development of three heating systems for L H D is under way. Fabrication of the actual heating systems will be started in 1995.

References [1] A. Iiyoshi, in Proc. IEEE 13th Symp. Fusion Engineering (SOFE), Knoxville, TN, USA, 2-6 October 1989, Vol. 2, IEEE, 1990, p. 1007.

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[2] Y. Takeiri, et al., Proc. First International Toki Conference on Plasma Physics and Controlled Nuclear Fusion, Toki City, Gifu, Japan, 4 7 December 1989, NIFS-PROC-3, National Institute for Fusion Science, Nagoya, 1990, p. 272. [3] K. Ohkubo, et al., Nucl. Fusion, in press. [4] T. Mutoh, et al., in Annual Report of the National Institute for Fusion Science (April 1991 March 1992), National Institute for Fusion Science, Nagoya, 1992, p. 61. [5] T.L. Owens, IEEE Trans. Plasma Sci. PS-14 (1986) 934. [6] T. Watari, et al., Phys. Fluids 21 (1978) 2076.

[7] Y. Takeiri, et al., in Proc. 16th Syrup. Fusion Technology, London, September 1990, Vol. 2, North-Holland, Amsterdam, 1991, p. 1012. [8] A. Ando, et al., in Sixth International Symposium on Production and Neutralization of Negative Ions and Beams, AIP Conference Proceedings No. 287, Upton, NY, USA, 1992, p. 339. [9] O. Kaneko, et al., in Fusion Technology 1992, Proc. 17th Syrup. on Fusion Technology, Rome, 14-18 September, 1992, Vol. 1, North-Holland, Amsterdam, 1993, p. 544.