Overview of Large Helical Device A. Komori, H. Yamada, O. Kaneko, N. Ohyabu, K. Kawahata, R. Sakamoto, S. Sakakibara, N. Ashikawa, P.C. deVries, M. Emoto, H. Funaba, M. Goto, K. Ida, H. Idei, K. Iked, S. Inagaki, N. Inoue, M. Isobe, S. Kado, S. Kubo, R. Kumazawa, S. Masuzaki, T. Minami, J. Miyazawa, T. Morisaki, S. Morita, S. Murakami, S. Muto, T. Mutoh, Y. Nagayama, Y. Nakamura, H. Nakanishi, K. Narihara, K. Nishimura, N. Noda, T. Kobuchi,, S. Ohdachi, Y. Oka, M. Osakabe, T. Ozaki, B.J. Peterson, K. Saito, A. Sagara, H. Sasao, M. Sasao, K. Sato, M. Sato, T. Seki, T. Shimozuma, M. Shoji, H. Suzuki, M. Takechi, Y. Takeiri, K. Tanaka, K. Toi, T. Tokuzawa, K. Tsumori, I. Yamada, S. Yamaguchi, S. Yamamoto, M. Yokoyama, Y. Yoshimura, K.Y. Watanabe, T. Watanabe, T. Watari, Y. Hamada, K. Itoh, K. Matsuoka, K. Murai, K. Ohkubo, I. Ohtake, M. Okamoto, S. Satoh, T. Satow, S. Sudo, S. Tanahashi, K. Yamazaki, O. Motojima and M. Fujiwara National Institute for Fusion Science Toki 509-5292, Japan Abstract: The Large Helical Device (LHD) experiments have started after a construction period of eight years, and two experimental campaigns were performed in 1998. The magnetic field was raised up to 2.75 T at a magnetic axis position of 3.6 m at the end of the second campaign. In the third campaign, started in July in 1999, the plasma production with ECH of 0.9 MW and auxiliary heating with NBI of 3.5 MW have achieved an electron temperature of 3.5 keV and an ion temperature of 2.4 keV. The maximum stored energy has reached 0.75 MJ with an averaged electron density of 7.7×1019 m-3 by a hydrogen pellet injection. The ICRF heating has sustained the plasma for longer than 2 s and the initial stored energy of the NBI target plasma has increased from 0.27 MJ to 0.335 MJ. The major characteristic of LHD plasma is the formation of temperature pedestal, which leads to some enhancement of energy confinement over the ISS95 scaling law. The confinement characteristic is gyro-Bohm and the maximum energy confinement has reached 0.28 s. LHD has also shown its high potentiality for steady state operation by realizing a 22 s discharge in the second campaign. 1. Introduction The Large Helical Device (LHD) offers a great opportunity to study currentless plasmas in the heliotron configuration. LHD is a large heliotron with a major radius R of 3.9 m and a minor radius a of 0.65 m [1-6]. Full steady state operation is expected, using superconducting coils in addition to a full helical divertor. Steady state plasma operation essential to magnetic fusion reactors is an inherent advantage of heliotrons over tokamaks and can be realized with the full helical divertor because it removes heat flux from the core plasma and controls impurity recycling. However, since the closed full helical divertor will not be ready in the near future, we plan to use a local island divertor (LID) for the LHD edge plasma control [7,8]. The LID is a closed divertor that uses an m/n = 1/1 island. Thus an LID coil system can be used to eliminate natural islands, which may be produced by error field, for example, due to ferromagnetic material located around LHD and the small misalignment of the helical and poloidal coils [7].

In the first and second campaigns, the magnetic field Bt was set at 1.5 T so as to accumulate operational experience the superconducting coils, and was raised up to about 2.5 T at the end of the second campaign. At this high field operation, the target plasma for NBI was found to be produced with the 84 GHz gyrotron in the region of Bt ≥ 2.2 T when a magnetic axis position Rax = 3.6 m, although an ECH beam does not cross the resonance area. Initiation of plasma discharge by NBI alone was also demonstrated at Bt = 1.5 T and Rax = 3.75 m. The plasma produced is comparable to the conventional one initiated by ECH and heated by NBI, so that the magnetic field scan in the region over 2.5 T becomes free in the experimental study. From the third experimental campaign, experiments was conducted at the magnetic field over 2.5 T with the heating power being increased gradually, aiming at producing higher temperature and density plasmas. The excitation test of the superconducting coils to the rated field of 3 T is planned after the third experimental campaign. Plasma production has started with ECH of 0.9 MW at Bt = 2.75 T and Rax = 3.6 m in the third campaign. NBI of 4.2 MW with negative ion sources optimized for the energy of 180 keV and ICRF of 0.9 MW with a pair of loop antennas for fast wave heating have been applied. This article is intended to present such plasma experiments in the third campaign as well as those in the first and second campaigns. 2. Experimental Apparatus LHD employs a pair of /m = 2/10 helical coils (liquid helium pool cooled), three sets of poloidal coils (super critical helium forced flow cooled) and nine pairs of buslines (two fluid helium forced flow cooled) [1]. The mass at the cryogenic temperature of 4.4 K is 822 t [6]. Transition to the superconducting state is achieved after three or four weeks' cooling of the coils. In addition to these superconducting coils, the LID coil system was installed in LHD, which consists of twenty island control copper coils, located around the upper and lower ports of the LHD cryostat. A clear m/n = 1/1 island is generated by this coil system at the ι/2π =1 surface in the edge region. Plasma heating devices using the ECH, NBI and ICRF techniques are employed in LHD experiments. We used ECH for plasma production and heating with gyrotrons at the characteristic frequencies of 84, 82.6 and 168 GHz. The R&D results were that the output power of the 84 GHz gyrotron is 0.45 MW for 10.2 s and 0.5 MW for 2 s. The ECH power was injected through two upper and one outer ports, and two transmission lines were installed in each port. The total ECH power is planned to be 1 MW with six gyrotrons in the third campaign. The elliptical Gaussian beam from the upper port was focused on the equatorial plane and the focal point can be changed radially and toroidally. Two negative ion based neutral beam injectors are the main heating devices and were aligned for tangential injection in opposite directions (co- and counter-injection) [9,10]. Each neutral beam injector was designed for injecting a 7.5MW hydrogen beam at the energy of 180 keV for 10 s by two ion sources. One of the two injectors was also designed to inject a 100 keV, 0.5 MW beam for up to 90 s. The ion sources for NBI were optimized for a low energy of 100 keV in the second campaign, considering that the target plasma density was low in the case of 1.5 T operation. The beam energy was increased up to 180keV, and the injection power will be raised up to several MW in the third campaign.

For the ICRF heating experiments, a folded wave guide antenna was installed after the second experimental campaign for ion Bernstein wave heating by which plasma production can be done independently of the magnetic field. The expected loading power is 0.5 MW for 1 s, which will be increased up to 1 MW for 10 s. The ICRF heating experiment with a power of 1 MW for 10 s also started in the third campaign, using a pair of loop antennas for fast wave heating. 3. Experimental results and Discussions 3.1. Magnetic configuration In helical systems, the existence of well-nested vacuum magnetic surfaces plays an essential role on plasma performance and it is possible to demonstrate the magnetic surfaces experimentally. Magnetic flux mapping was done under steady state conditions, using the fluorescence method [5,11]. In the measurements, an electron gun with a LaB6 cathode and a fluorescent mesh were used. The diameter of the electron beam was limited to less than 3 mm. The fluorescent mesh of 1m × 2 m was located at the toroidal plane, the cross-section of which is elongated vertically. The magnetic surfaces were measured by taking pictures of the beam images on the fluorescent mesh while changing the radial position of the electron gun. The pictures were taken with a charge coupled device (CCD) camera, which was aligned in parallel with the tangential port for NBI. The mapping at reduced magnetic fields ranging up to 0.25 T, performed just after the first campaign, showed clearly that there are closed, nested magnetic surfaces, in good agreement with the theoretical calculation [5,11]. The mapping at Bt = 2.75 T and Rax = 3.6 m was also carried out after the second campaign. There are an m/n = 1/1 island with a maximum width of about 8 cm and 2/1 islands with a maximum width of about 5 cm. The cause of such large islands is not clear at this stage, and there are a variety of possibilities, for example, the existence of ferromagnetic material located around LHD and an error field arising from coil current feeders, in addition to the large misalignment of the coils. An error field due to the 2 mm misalignment of the coils, which is the maximum tolerance permitted in the coil specification, forms the m/n = 1/1 island with a maximum width of only about 4 cm. The size of the islands can be, however, minimized by the LID coil system. Both 1/1 and 2/1 islands are almost eliminated by the LID coil system simultaneously, and the more accurate magnetic surfaces are realized, than those formed with the 2

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Fig. 1 Dependence of Wp/Wp(ILID = 0) on ILID. Here ILID is a current of island control coils, normalized by the current, necessary for generating an m/n = 1/1 island of about 15 cm width.

mm misalignment of the coils. The plasma performance was examined in the magnetic configuration with the minimized islands in the third campaign. A comparison between NBI plasma discharges with and without the islands was performed with the NBI power of 3 MW at Bt = 2.75 T and Rax = 3.6 m, and plasma parameters were changed little. Figure 1 shows the dependence of stored energy Wp on a normalized current ILHD. Here, ILHD is a current of the island control coils, normalized by the standard current that is required for generating the m/n = 1/1 island of about 15 cm width when the island due to error field does not exist. The stored energy Wp is found to decrease with ILHD, and this is attributed to the change of the plasma volume that becomes small with an increase in ILHD, i.e. the island volume. Thus the small effect of the minimized islands on Wp is also explained by this change of the plasma volume. The result with the minimized magnetic islands is plotted at ILHD = -0.22 in Fig. 1. Accordingly we usually do not use the magnetic configuration with the minimized islands. The existence of the islands can be recognized in the electron temperature Te profile measured by Thomson scattering, as shown in Fig. 2. The m/n = 1/1 island is clearly shown at the ι /2π = 1 surface positions forming the flat profiles when ILHD = 0.68. Although the island is not recognized at ILHD = 0, as shown in Fig. 2(a), the small 1/1 island is sometimes observed due to the error field. 3.2. Wall conditioning Baking the plasma vacuum vessel at temperatures of around 95 ˚C and ECR discharge cleaning with hydrogen or helium gases were performed after the first plasma production in the first campaign. A 2.45 GHz ECR system of 5 kW power is used for the ECR discharge cleaning, with the standard magnetic configuration of 0.0875 T [5]. The main desorbed gases were hydrogen, carbon monoxide and water, and the total amount of evacuated oxygen was about 100 molecular layers, i.e. about 50g. To enhance wall conditioning, the toroidally distributed Ti gettering was done once or twice a day, which covers 20-30% of the plasma facing wall surface. In the second campaign, after the plasma vacuum vessel was baked for 240 h, helium glow discharge cleaning with a power of 5 kW,

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Fig. 2. Electron temperature profiles with and without the LID, together with the ι/2π=profile. ILID = 0.68.

partly combined with helium ECR discharge cleaning, was curried out, including the conditioning of the NBI drift tubes. This resulted in good wall conditioning, and more or less the same result as in the first campaign was obtained. Titanium gettering was also done in the initial phase of the second campaign. The effective Z number Zeff of hydrogen discharge was about two at the end of the second campaign, and the main impurity was oxygen.

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Fig. 3. Total pressure versus partial pressures of water and carbon monoxide. Blue circles were obtained on 15 July, while red circles were on 3 September.

Fig. 4. Comparison between spectra of impurities before (the upper figure) and after (the lower figure) the installation of divertor carbon tiles.

In the third campaign, the plasma vacuum vessel was baked for 770 h and the helium glow discharge cleaning with a power of 5 kW was carried out for 1,120 h, until the end of September. Figure 3 shows the total pressure versus partial pressures measured with a mass analyzer, and represents all data taken before, during and after the plasma discharges. The impurity pressures taken on 3 September are found to decrease clearly, compared with those on 15 July at the same total pressures. The later was measured on the early stage of the third campaign, while former was after the long wall conditioning. The partial pressures are almost the same as those obtained in the end of the second campaign. Thus LHD is already in good wall condition. Helium ECR discharge cleaning with the power of 16 kW was also performed for 9 h, and its effect on the plasma performance was compared with that after the glow discharge cleaning. As a result it became clear that the ECR discharge is not effective for cleaning in LHD. Titanium gettering will be done in October to make sure of its effect of impurity suppression. Carbon divertor plates were installed at the divertor leg for the protection of the plasma vacuum vessel before the third campaign. The carbon plates are located on the copper base plates, cooled by water, and the copper base plates are fixed on the stainless steel cooling pipes. The effect of the carbon plates on the plasma performance is not clear at this stage, but impurities desorbed from the stainless steel wall were reduced extremely, as shown in Fig. 4. Thus it is difficult to use the iron emission line for the spectroscopic measurement of the ion temperature Ti [12]. 3.3. Plasma performance In the heating experiment with ECH alone, two 84 GHz, one 82.6 GHz and three 168 GHz gyrotrons were used for producing a plasma. The total input power of ECH was around 0.9 MW for about 0.1 s, although the duration of each gyrotron is longer than this value. Although the ECH beams from the upper ports do not cross the resonance area at Bt = 2.75 T and Rax = 3.6 m, there are the resonance areas at ρ = -0.2∼ -0.3 for the ECH beams from the outer port. The averaged electron density n−e and Wp were measured to be 3∼ 5×1018 m-3 and 0.07 MJ, respectively. The electron temperature reaches 3.5 keV at the plasma center.

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Fig. 5. Radial electron temperature, ion temperature and electron density profiles.

The auxiliary heating with NBI makes new records day by day with an increase in a power of NBI. The profiles of Ti and Te are shown in Fig. 5, together with that of n−e The density profiles of NBI plasmas are almost flat, while the temperature profiles are triangular like. The Ti profile agrees well with that of Te. In Fig. 5(a), it should be noted the edge temperatures are fairly high, suggesting the formation of temperature pedestal. The plasma parameters attained up to now are summarized as follows: (1) Te of 3.3 keV, Ti of 2.4 keV, the energy confinement time τΕ of 0.16 s and n−e of 1.5×1019

m-3 at the absorbed power Pabs of 2.6 MW; (2) τE of 0.28 s, Te of 2.1 keV and n−e of 3.4×1019 m-3 at Pabs of 1.7 MW; (3) nτT = 1.6×1019 keVm-3s; (4) the maximum Wp = 0.75 MJ; (5) <β> = 1.0%; and (6) the maximum n−e = 7.7×1019 m-3. The maximum Wp and n−e were obtained with a hydrogen pellet injector, which can inject a series of 5 pellets at intervals of 1 ms ∼ =1 h. Although the experiment was usually performed at Rax = 3.6 m, data were also taken at Rax = 3.9 m and 3.75 m. The comparison among these data indicates that the best plasma performance is achieved at Rax = 3.6 m, as expected from the calculation of particle orbits. The time evolution of plasma parameters, obtained in the hydrogen puffing NBI discharges at Bt = 1.5 T and Rax = 3.75 m mainly performed in the second campaign, showed that n−e decreases significantly and the diverted plasma flux increases during this pumping phase. The decrease in n−e occurs when the edge temperatures are high, typically, 200∼ 400 eV. Since LHD employs metal walls, that is, stainless steel walls, it is considered that the metal walls strongly pump the hydrogen plasma and that the global recycling coefficient in LHD is less than unity. The increase in NBI power results in a shorter decay time of the density drop. By contrast, n−e is found to be controlled by gas puffing during the discharge at Bt = 2.75 T and Rax = 3.6 m. One reason of this difference is considered to be that the rate of plasma loss by radial diffusion is decreased because of the high magnetic field. The ICRF heating experiment was carried out with a pair of loop antennas [13]. The input power of ICRF was 0.5∼ 0.9 MW with the RF frequency of 38.47 MHz for a pulse duration of about 2 s. The target plasma was produced by ECH or NBI at Bt = 2.75 T and Rax = 3.6

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m. The helium gas puffing was used, containing hydrogen gas as a minority. The target plasma density was 0.4∼ 1.4×1019 m-3. The coupling resistance of plasma was 5∼ 8 Ohm when the distance between the LCFS and loop antenna front was 3∼ 5 cm, while it was 0.8 Ohm in vacuum. When the target plasma was produced by ECH alone, the plasma was sustained by ICRF for about 2 s. The maximum stored energy Wp is obtained to be 51.4 kJ with the ICRF power of 0.5 MW at H/(He + H) = 50%. The maximum increment of Wp is an increasing function of the ICRF power in this power range, so that much more increase in Wp is expected with the ICRF power. Thus the ICRF heating has a good heating efficiency comparable to that of the ECH heating in this power range, since the power of ECH was also around 0.5∼ 0.9 MW. Figure 6 shows the temporal evolution of plasma parameters during the ICRF heating, applied to the NBI discharge. The target plasma was produced with one NB injector of 1.5∼ 1.6 MW, and Wp is increased from 0.27 MJ to 0.335 MJ. These results can be explained by the one dimensional wave analysis calculation based on the slab plasma model. As mentioned before, we can observe a temperature pedestal at the edge of the Te profile. The width of the pedestal is 4∼ 6 cm in the short minor radius direction and the value of Te

ped (at ρ = 0.9) is very high, considering that it is 30∼ 40% of the central temperature and 70-80% of the average temperature. The observed maximum Te

ped is higher than 1 keV. Thus Teped determines Wp and hence

τE. Formation of pedestal takes place naturally during the rising phase in the discharge. A shrunken plasma, generated by ECH, expands with the NBI heating and then the hot plasma region reaches the LCFS, increasing Te

ped and the divertor flux during 10∼ 20 ms. The pedestal is normally located just inside the LCFS. But there are a few observations to suggest that the high Te region is around the ι =/2π== 1 surface. The cold plasma usually exists outside the LCFS and n−e at ρ = 1.0 is almost equal to that at the plasma center. These Te and n−e profiles were found to depend on the magnetic configuration. Figure 7 shows the time evolution of radiation profile measured with a bolometer. The radiation power is calculated to be about 40% of deposition power during the steady state portion of discharge. The radiation profile is very hollow for entire discharge, and asymmetric, stronger on the outboard side. The emission peak is clearly located outside the LCFS, corresponding to the pedestal location. To study the role of the ergodic layer in the formation of temperature pedestal, we planed the experiments of changing a plasma radius with a limiter [14]. As shown in Fig. 8(a), the limiter is movable and can be inserted into the LCFS to cut the edge plasma in the ergodic layer. The limiter head is made of carbon plates that cover a stainless steel plate. When the limiter is inserted into the plasma, n−e becomes high, together with a decrease in Wp, i.e. a decrease in the temperatures. These are attributed to the large amount of gases desorbed from the limiter when the plasma strikes the limiter. Figure 8(b) shows that the pedestal is formed even if the limiter is inserted into the LCFS and the

Fig. 7. Emissivity profile evolution.

plasma in the ergodic layer is eliminated. These suggest that the ergodic layer and fairly strong pumping in the edge region are not necessary for the formation of the temperature pedestal. However, further investigations are necessary for clarifying the reason why the pedestal is formed. Confinement characteristics of the LHD plasmas were examined using the NBI discharges. Parameter regimes studied here cover Bt of 1.5∼ 2.75 T, n−e of 1.0∼ 4.9×1019 m−3 and Pabs of 0.75∼ 3.2 MW. Although LHD has a variety of magnetic configurations, behaviors only in the standard configuration, characterized by Rax of 3.75m and the toroidal averaged ellipticity of 1, are studied. The database in this study consists of 68 discharges in quasi steady state. Comparison of LHD data with the available scaling laws, i.e. the International Stellarator Scaling 95 (ISS95) [15], the Lackner--Gottardi (L-G) scaling [16] and the LHD scaling [17], indicates systematic enhancement of confinement. Figure 9(a) depicts the energy confinement time τE and the ISS95 scaling law, τE

ISS95 = 0.079 a2.21 R0.65 Bt

0.83 n−e0.51 Pabs

-0.59 (ι /2π)2/30.4. The energy confinement time τE was estimated using Wp

measured with the diamagnetic loops. The response of the diamagnetic loops has been calculated by the 3-D magnetic field analysis with 3-D finite-β equilibrium. The NBI deposition profiles were evaluated using a Monte Carlo simulation code, MCNBI [18], and the ECH deposition power was derived from the water load measurement considering losses due to mirrors and bends in the wave guides. Enhancement factors are 1.18 + 0.15 for the ISS95 scaling, 1.15 + 0.13 for the L-G scaling and 1.44 + 0.16 for the LHD scaling. In particular, a factor of two enhancement has been observed with the comparison of the scaling expression based on the medium size heliotrons (Heliotron E, ATF and CHS). The expression of this scaling is almost dimensionally correct and has very similar power dependence to the ISS95 scaling. Regression analysis of the LHD data alone gives τE = 0.05 Bt

0.71 n−e0.61

Pabs-0.61 which can be written into τE = τΕ

B ρ∗ -0.83 β0.08 ν∗ 0.01 q0.01 L2.82, where L with the dimension of length is selected to have correct dimensions. The unified size scaling is not simple, however, the dependence on density, magnetic field and absorbed power is robust and suggests gyro-Bohm type

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characteristics [19] in LHD as well. The major characteristic of LHD plasma, different from the past medium size heliotrons, is the formation of pedestal. This is contrast to the fact that the plasma pressure in the medium size heliotrons decreases asymptotically to zero towards the LCFS. We assumed that the confinement region is divided into the core and pedestal regions, and the core value should be compared with the scaling. The core value is defined by the subtraction of the pedestal at ρ = 0.9 from the total value. A statistical analysis of the data from the combination of the core confinement in LHD and the whole confinement of the medium size heliotrons gives the fitting expression of τE

scl1 = 0.04 a2.09 R0.75 Bt0.80

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present understanding that the confinement of currentless helical plasmas is gyro-Bohm type. The energy confinement time derived from the value obtained by subtracting the pedestal at ρ== 0.9 from Wp, were found to agree quite well with this scaling, as shown in Fig. 9(b). In conclusion, the enhancement of confinement in LHD can be attributed to the significant formation of pedestal and the confinement characteristic is gyro-Bohm. Local transport analysis has been also conducted to clarify confinement enhancement in LHD. Since ρ* is the key parameter, comparison with CHS and LHD is informative. Comparison of two discharges in CHS and LHD with the same collisionality and different ρ* by a factor of two indicates that the electron heat conductivity scales well with ρ* in the core region. Since the temperature asymptotically decreases to zero towards the edge in CHS, ρ* decreases significantly at the edge. In contrast, ρ* in LHD decreases much more gradually due to high edge temperature. This trend predicts that the heat conduction should be improved at the edge in CHS, compared with LHD. However, the experimental observation has shown the opposite trend that the electron heat conductivity deteriorates rapidly towards the edge in CHS and is flat in LHD. The argument in global confinement improvement with related to the formation of pedestal is supported by this local transport analysis. Regarding to comparison with the neoclassical theory, the deterioration factor is suppressed up to around 3 in LHD although it has the order of 10-102 in CHS. However, the anomalous transport

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is still dominant in NBI heated plasmas in LHD, although the collisionality enters the collisionless regime where the helical ripple transport plays an essential role. In contrast to the electron transport, the ion transport is close to the neoclassical theory. The values of both heat conduction coefficients for ions and electrons are close to each other in LHD. The study of MHD instabilities has just started with magnetic probes [20] and soft X-ray detectors. The instabilities whose frequency was lower than 50 kHz were measured with the magnetic probes, and preliminary analysis has been done [20]. 3.4. Long pulse discharge Quasi-steady-state operation with a duration of 22 s was attained with NBI at Rax = 3.75 m and Bt = 1.5 T in the second campaign. A 0.67 MW of 66 keV hydrogen beam was injected into the target plasma. Except for a small amount of helium gas puffing at the start of discharge, no additional gas was puffed during the discharge. Thus n−e was as low as 0.3×1019 m-3, although 40% of the beam was absorbed. In this case, the duration of the discharge was determined by the NBI conditions, and the effort will be continued to extend the duration of discharge at higher input power. When we tried to increase n−e, slow oscillations of plasma parameters were observed or the discharge was terminated by collapse at Rax = 3.75 m and Bt = 1.5 T in the second campaign. The TV image of the plasma showed that the radius of a hot core repeats a series of expansion and contraction like “breathing”. The threshold density was found to be around 1×1019 m-3. In the third campaign, however, the breathing phenomenon was not observed even if n−e exceeds the threshold. This reason is not clear at this stage. Figure 10 shows the stable discharge with the duration of about 10 s. The electron density n−e is about 1×1019 m-3 at t = 10 s, and Ti is 2.1 keV at t = 5.45 s and R ∼ 3.8 m. 4. Summary After challenging technology R&D over a period of eight years, the construction of LHD has been completed. Auxiliary heating experiments with NBI of 2.6 MW have achieved an electron temperature of 3.3 keV and an ion temperature of 2.4 keV with an averaged electron density of 1.5×1019 m-3 at a magnetic field of 2.75 T and magnetic axis position of 3.75 m. The maximum stored

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Prad

Fig. 10. Temporal behaviors of stored energy, electron density and radiation power, observed in a 10 s discharge.

energy has reached 0.75 MJ. The energy confinement time has shown some enhancement over the present scaling law, which was derived from existing medium size helical devices, and the maximum value has reached 0.28 s. LHD has also shown its high potentiality for steady state operation by realizing a 22 s discharge. The unique plasma initializing technique with NBI has been demonstrated, which is possible only if a vacuum magnetic field confines high energy beams. The rest of the third campaign is focusing on realizing higher NBI power to achieve the target and study high plasma performance. Optimization of the magnetic structure, investigation of divertor functions and study of MHD characteristics will be also studied. Acknowledgements The authors would like to acknowledge the continuing encouragement of A. Iiyoshi. References [1] A. Iiyoshi et al., Fusion Technol., 17, 169 (1990) . [2] O. Motojima et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (IAEA, Vienna, 1991) Vol. 3, pp. 513. [3] K. Yamazaki et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (IAEA, Vienna, 1991) Vol. 2, pp. 709. [4] M. Fujiwara et al., Plasma Phys. Control. Fusion, 39, A261 (1997). [5] A. Iiyoshi et al., Plasma Physics and Controlled Nuclear Fusion Research 1998 (IAEA, Vienna, 1999) IAEA-CN-69/OV1/4 (to be published). [6] O. Motojima et al., Plasma Physics and Controlled Nuclear Fusion Research 1998 (IAEA, Vienna, 1999) IAEA-CN-69/FT2/1 (to be published). [7] A. Komori et al., Plasma Physics and Controlled Nuclear Fusion Research 1994 (IAEA, Vienna, 1995) Vol. 2, pp. 773. [8] A. Komori et al., Fusion Energy 1996 (IAEA, Vienna, 1997) Vol. 2, pp. 3. [9] Y. Takeiri et al., Rev. Sci. Instrum., 66, 2541 (1995). [10] O. Kaneko et al., Fusion Energy 1996 (IAEA, Vienna, 1997) Vol. 3, pp. 539. [11] O. Motojima et al., Phys. Plasmas, 6, 1843 (1999). [12] S. Morita et al., this conference. [13] T. Seki et al., this conference. [14] K. Nishimura et al., this conference. [15] U. Stroth et al., Nucl. Fusion, 36, 1063 (1996). [16] K. Lackner and N.A.O. Gottardi , Nucl. Fusion, 30, 767 (1990). [17] S. Sudo et al., Nucl. Fusion, 30, 11(1990). [18] S. Murakami et al., Trans. Fusion Technol., 27, 256 (1995). [19] M. Murakami, Plasma Physics and Controlled Nuclear Fusion Research 1992 (IAEA, Vienna, 1993) Vol. 2, pp. 391. [20] S. Sakakibara, this conference.