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Okuno, Kiyoshi; Nakajima, Hideo; Sugimoto, Makoto; Isono, Takaaki
Fusion Engineering and Design, 81(20-22), p.2341 - 2349, 2006/11
Times Cited Count：2 Percentile：17.18(Nuclear Science & Technology)Japan is expected to make a major contribution to the procurement of the ITER superconducting magnet system. The ITER magnet system consists of 18 TF coils, one CS with six modules and six PF coils. The TF coils and CS have major features in unprecedented size of the magnets and structures and operations at 11 to 13 T high fields that require Nb
Sn superconductor. Because of these features, significant efforts were made towards developing superconducting magnet technology to a level that will allow the ITER magnets to be built with confidence. The construction and testing of the CS and TF model coils have therefore been performed during the ITER Engineering Design Activity (EDA), and all ITER RD goals have been achieved. Based on these achievements, further development activities are now being performed at JAERI in tight collaboration with industry for the preparation of the ITER construction. The activities include analytic studies on design improvement and optimization, manufacturing studies to identify the detailed fabrication processes and tools, and manufacturing demonstrations on full-scale structural components (several tens of tons) such as made of new cryogenic materials, JJ1 and strengthened 316LN. Results from these activities will provide firm technical basis to achieve required performance of the magnets while maintaining both project schedule and cost and to reduce technical risks that may happen during the manufacturing phase.
Sato, Satoshi; Iida, Hiromasa; Nishitani, Takeo
Fusion Engineering and Design, 81(23-24), p.2767 - 2772, 2006/11
Times Cited Count：14 Percentile：68.12(Nuclear Science & Technology)no abstracts in English
Kasugai, Atsushi; Minami, Ryutaro; Takahashi, Koji; Kobayashi, Noriyuki; Sakamoto, Keishi
Fusion Engineering and Design, 81(23-24), p.2791 - 2796, 2006/11
Times Cited Count：9 Percentile：53.38(Nuclear Science & Technology)In JAEA (Japan Atomic Energy Agency, formerly JAERI), development of 170 GHz, 1 MW, CW gyrotron for ITER has been carried out. Key technologies for ITER gyrotron such as a diamond window, a depressed collector for high efficiency operation and a stable operation at 170 GHz/1 MW with higher mode TE31,8 have been developed. By integration of these key technologies, gyrotron performance of 0.5 MW/ 100 sec and 0.9 MW/ 9.2 sec were demonstrated. Hence, next target is a demonstration of long pulse operation. One of the issues which prevent the pulse extension is large beam current decrease due to so called the emission cooling of a cathode. During the operation, the oscillation mode shift from TE31,8 to TE30,8 was caused by the current decrease. Then, the magnetic field of the cavity should be increased to avoid the downshift of the oscillation mode, however, the efficiency decreases and the parasitic oscillation appears in a quasi-optical mode converter. To suppress the beam current decrease and to demonstrate the long pulse operation of the high power gyrotron, pre-programming control of the cathode heater power was applied and the long pulse experiment was carried out to sustain the beam current. As a result, stable electron beam of 1000 s, which is required for ITER operation, was demonstrated without oscillation, and pre-programming control directed the effectiveness for constant beam current. Moreover, in the experiment of the long pulse oscillation with oscillation, the pre-programming control suppressed the beam current decrease. Up to now, stable long pulse operation of 
8 minutes with 0.2 MW output power was obtained. The output energy of the oscillation is maximum value in the 170GHz ITER gyrotron. Since overheating due to stray radiation inside the gyrotron limit the pulse extension, long pulse operation with high power output will be achieved by enhancement of the cooling and reduction of stray radiation due to modification of a built-in mode converter.
Kashiwagi, Mieko; Hanada, Masaya; Yamana, Takashi*; Inoue, Takashi; Imai, Tsuyoshi*; Taniguchi, Masaki; Watanabe, Kazuhiro
Fusion Engineering and Design, 81(23-24), p.2863 - 2869, 2006/11
Times Cited Count：4 Percentile：30.68(Nuclear Science & Technology)Plasma neutralizer is one of key components to achieve the required system efficiency (
50 %) for a negative-ion based neutral beam (N-NB) system in a fusion power plant. In the plasma neutralizer, highly ionized plasma is required at lower pressure, e.g., ionization degrees of 30 % at 
0.08 Pa for 1 MeV negative ions. In such low pressure, mean free path of fast electron that contributes to ionizations becomes longer than the neutralizer's dimensions. Therefore, suppression of fast electron leakage from large openings that are beam entrance and exit is a crucial issue to realize plasma neutralizers. To suppress the fast electron leakage from the openings, authors propose a shield field, which is a weak transverse magnetic field of only 30 Gauss applied locally around the opening. The shield field are numerically examined and designed by using a three dimensional particle orbit code. In the experimental studies, this weak shield field is applied at the openings (diam. = 20 cm) of an arc discharge driven plasma neutralizer (length = 200 cm, diam. = 60 cm). The plasma parameters inside and outside of the opening were measured by a Langmuir probe. The electron energy distribution function (EEDF) showed that considerable fast electrons, which were leaked from the opening, were suppressed successfully by the weak shield field of 30 Gauss. Thus the leaking fast electrons were repelled into the neutralizer to deposit their energy for the plasma production. At a result, the plasma production efficiency (plasma density / arc power) was improved by a factor of 1.5 at 0.08 Pa.
Sato, Masayasu; Nishio, Satoshi; Tobita, Kenji; Inoue, Takashi
Fusion Engineering and Design, 81(23-24), p.2725 - 2731, 2006/11
Times Cited Count：1 Percentile：9.98(Nuclear Science & Technology)no abstracts in English
Hayashi, Takao; Tobita, Kenji; Nishio, Satoshi; Sato, Satoshi; Nishitani, Takeo; Yamauchi, Michinori
Fusion Engineering and Design, 81(23-24), p.2779 - 2784, 2006/11
Times Cited Count：2 Percentile：17.18(Nuclear Science & Technology)We have studied the possibility of tritium self-sufficiency with the outboard blanket only in low aspect ratio tokamak reactor to simplify the inboard structure in high magnetic field. The tritium breeding ratio (TBR) of the outboard blanket increases by applying the inboard reflector working as a neutron multiplier such as lead and beryllium. Assuming that the coverage of the breeder zone to plasma facing components is 0.78, the requirement of local TBR is above 1.35. The local TBR calculated with both the inboard reflector of lead and the outboard blanket using beryllium multiplier and Li
O breeder is larger than 1.35 for aspect ratios smaller than 2.9. These results indicate there is a solution for tritium self-sufficiency using the outboard breeding blanket only in low-A tokamak reactor when the appropriate inboard reflector and outboard blanket are adopted. In case both inboard and outboard blankets are applied, the total TBR with the combination of LiO and Be is more than 1.5. However, Be
Ti and LiTiO are recommended from the viewpoint of safety, and the TBR with the combination is slightly larger than 1.35 for aspect ratios from 2 to 4.
Fukumoto, Naoyuki*; Ogawa, Hiroaki; Nagata, Masayoshi*; Uyama, Tadao*; Shibata, Takatoshi; Kashiwa, Yoshitoshi; Suzuki, Sadaaki; Kusama, Yoshinori; JFT-2M Group
Fusion Engineering and Design, 81(23-24), p.2849 - 2857, 2006/11
Times Cited Count：7 Percentile：46.04(Nuclear Science & Technology)no abstracts in English
Li, Z.*; Tanaka, Teruya*; Muroga, Takeo*; Sato, Satoshi; Nishitani, Takeo
Fusion Engineering and Design, 81(23-24), p.2893 - 2897, 2006/11
Times Cited Count：1 Percentile：9.98(Nuclear Science & Technology)no abstracts in English
Nakamura, Yukiharu; Tobita, Kenji; Takei, Nahoko; Takase, Yuichi*; Fukuyama, Atsushi*; Nishio, Satoshi; Sato, Masayasu; Jardin, S. C.*
no journal, ,
no abstracts in English
Nishimura, Arata*; Hishinuma, Yoshimitsu*; Seo, Kazutaka*; Tanaka, Teruya*; Muroga, Takeo*; Nishijima, Shigehiro*; Katagiri, Kazumune*; Takeuchi, Takao*; Shindo, Yasuhide*; Ochiai, Kentaro; et al.
no journal, ,
A fusion device which creates burning plasma will be equipped with a superconducting magnet system to provide strong magnetic field and maintain the burning plasma. The fusion device also will have plasma heating devices such as neutral beam injectors and electron cyclotron systems. Since these systems need several ports to carry in the energy into plasma, the fusion device has large ports connecting to the systems locates in outside of cryostat. Through these ports, D-T neutron will come out of the burning plasma and damage the surrounding materials. The superconducting magnets also will be irradiated by the streaming neutron. To investigate mechanisms of degradation of superconducting properties, and to construct database of irradiation effect on superconducting magnet materials, a cryogenic target system has been install in Fusion Neutronics Source (FNS) at Japan Atomic Energy Agency (JAEA). The irradiation tests with D-T neutron have been carried out three times and some irradiation effects on superconducting magnet materials are clarified. In this paper, the present status of the cryogenic target system and some irradiation test results will be summarized and presented.
Nishio, Satoshi
no journal, ,
The CS coil system is an obstacle to a low cost construction. On the other hand, an operational flexibility (or capability) of CS-less tokamaks is restricted somewhat. This somewhat should be cared about. To be acceptable or not. To be compensatable or not. That is question.Even though the CS coil system is discarded from a tokamak device, plasma initiation and succeeding current ramp up are not any longer a critical issue of possible or impossible. The concerns will be shifted to how skillfully or how fast issues. More important issue is a plasma equilibrium forming. It is difficult to form an elongated plasma shape with high triagularity for CS-less poloidal field (PF) coils arrangement. Because there is no coil to push in the plasma column in a torus inner region. Focusing on the plasma shape being with reference to a plasma internal inductance, the operational parameter region is illustrated in elongation-triagularity-internal inductance space. Simultaneously and furthermore, allowable plasma elongation is discussed in terms of passive shell positioning, stability margin and plasma aspect ratio. Maximum achievable and minimum required triagularity are discussed.
