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Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Hirakawa, Yasushi; Furukawa, Tomohiro; Hoashi, Eiji*; Fukada, Satoshi*; Suzuki, Akihiro*; Yagi, Juro*; Tsuji, Yoshiyuki*; et al.
Proceedings of Plasma Conference 2014 (PLASMA 2014) (CD-ROM), 2 Pages, 2014/11
In the IFMIF/EVEDA (International Fusion Materials Irradiation Facility/ Engineering Validation and Engineering Design Activity), the validation tests of the EVEDA lithium test loop with the world's highest flow rate of 3000 L/min was succeeded in generating a 100 mm-wide and 25 mm-thick free-surface lithium flow steadily under the IFMIF operation condition of a high-speed of 15 m/s at 250
C in a vacuum of 10 
Pa. Some excellent results of the recent engineering validations including lithium purification, lithium safety, and remote handling technique were obtained, and the engineering design of lithium facility was also evaluated. These results will advance greatly the development of an accelerator-based neutron source to simulate the fusion reactor materials irradiation environment as an important key technology for the development of fusion reactor materials.
Tojo, Hiroshi; Hatae, Takaki; Sakuma, Takeshi; Hamano, Takashi; Itami, Kiyoshi; Suito, Satoshi*; Araki, Takashi*; Iwamoto, Kohei*; Takeda, Yuya*
no journal, ,
Saito, Makiko; Maruyama, Takahito; Ueno, Kenichi; Takeda, Nobukazu; Kakudate, Satoshi
no journal, ,
In ITER, after plasma operation, The Blanket Remote Handling System (BRHS) will be installed in the vacuum vessel and it will remove and install the shield blanket module. BRHS will undergo hands-on maintenance in the maintenance area after the exchange of the shield blanket module. Since BRHS will be contaminated the radioactive dust in the vacuum vessel, the workers will be exposed by radioactive dust. In this study, potential contaminated areas and their respective dose rates from the BRHS using MCNP5 code to assess the exposure of maintenance workers. The assessment was performed using 3 types of equipment, vehicle manipulator, combination of cable handling and rail support, and sliding beam, which are installed in vacuum vessel or port. The dose calculations used the nuclides Ta-182 and W-181 and the dose was calculated from each of the 20 points spaced evenly around the equipment. As a result, there are some local points with high dose rates, which are exceed the target of acceptable dose limit for hands-on work in ITER (5 
Sv/h) in vehicle manipulator and combination of cable handling and rail support. To decrease the dose rate, lead blocks were used for shielding and as a result, the dose rate decreased to around 2.5 Sv/h using 5 mm and 10 mm lead shielding.
Imazawa, Ryota; Ono, Takehiro; Kawano, Yasunori; Itami, Kiyoshi
no journal, ,
no abstracts in English
Ogawa, Hiroaki; Kitazawa, Sin-iti; Sugie, Tatsuo; Katsunuma, Atsushi*; Kitazawa, Daisuke*; Omori, Keisuke*; Itami, Kiyoshi
no journal, ,
no abstracts in English
Sugie, Tatsuo; Takeuchi, Masaki; Ishikawa, Masao; Shimada, Takahiko; Katsunuma, Atsushi*; Kitazawa, Daisuke*; Omori, Keisuke*; Itami, Kiyoshi
no journal, ,
no abstracts in English
Sato, Satoshi; Iida, Hiromasa*; Tanigawa, Hisashi; Hirose, Takanori; Ochiai, Kentaro; Konno, Chikara; Enoeda, Mikio
no journal, ,
Shielding analyses have been performed on the ITER/TBM and its port by using Monte Carlo radiation transport calculation code MCNP and activation calculation code ACT-4. CAD data were automatically converted to MCNP data on the TBM, shield, pipes and bio-shield by the CAD/MCNP conversion code, and their MCNP data have been inserted to the ITER 40 degree model. We evaluated the effective dose rate in operation, nuclear heating, tritium production rate, effective dose rate after shutdown and induced activity.
Takeuchi, Masaki; Sugie, Tatsuo; Takeyama, Shigeharu; Itami, Kiyoshi
no journal, ,
no abstracts in English
Yatsuka, Eiichi; Hatae, Takaki; Takeyama, Shigeharu; Bassan, M.*; Vayakis, G.*; Walsh, M.*; Itami, Kiyoshi
no journal, ,
In ITER, 10
pulses of 5-J laser will be injected during its operation for 20 years. It was concerned that the surface damage was induced due to high incident laser energy density if conventional beam dump would be used. JAEA has proposed and modified the detailed shape of a new type beam dump in which the laser energy is reflected numerous times and absorbed gradually due to low absorptance of the S-polarized light. The new type beam dump withstands 10 pulses injection while conventional beam dump withstands 10
pulses (100 seconds) injection. In order to verify this design for drastically extending the beam dump life time, laser resistance test for molybdenum that is the beam dump material was carried out. As expected in design, beam energy absorption density is dominant factor for laser induced damage. Moreover, it was confirmed that interlaminating the bent sheets and shims and their diffuse bonding enable to manufacture complicated shape of the beam dump as designed.
Shimada, Takahiko; Ishikawa, Masao; Yatsuka, Eiichi; Hatae, Takaki; Itami, Kiyoshi
no journal, ,
Neutronic analysis has been carried out for the detailed design of Edge Thomson Scattering System (ETS) which is procured by JAEA in the ITER project. ETS measures the profile of electron temperature and density in the edge region of the plasma and consists of the laser injection system and the optical collection system installed in the EQ port plug. The optical collection system consists of many mirrors. High nuclear heating of mirrors could distorts them and deteriorates measurement accuracy. In order to reduce nuclear heating of optical mirrors, the effect of materials on nuclear heating of the optical mirrors has been evaluated. As a result, nuclear heating rate can be reduced about 25% by replacing the material of the first mirror (plasma facing mirror) from molybdenum to SUS316. Also it was found that nuclear heating rate of other mirrors can decrease about 65% by replacing materials from SUS316 to Aluminium. As described above, the appropriate material for each optical mirrors can be selected by nutronic analysis.
Kamiya, Kensaku; Honda, Mitsuru; Urano, Hajime; Yoshida, Maiko; Kamada, Yutaka
no journal, ,
In this study, we perform to identify the boundary condition for the toroidal plasma flow, which is known to play an important role for an improved plasma stability and/or confinement, to improve the prediction in the plasma core region by a code, experimentally. Boundary condition of toroidal plasma flow imposed at the separatrix in JT-60U ELMy H-mode plasmas has been identified, comparing between co- and counter-NBI discharges. Toroidal plasma flow at the separatrix can become non-zero value, varying with momentum input direction, but being unaffected by the ELM event. The results indicate that the viscosity damping due to SOL flow through ion-neutral collision may not affect strongly ion parallel momentum transfer.
Nakano, Tomohide; Shumack, A.*; Maggi, C. F.*; Reinke, M.*; Lawson, K.*; P
tterich, T.*; Brezinsek, S.*; Lipschultz, B.*; Matthews, G. F.*; Chernyshova, M.*; et al.
no journal, ,
In an European tokamak device, JET, an existing high-resolution X-ray crystal spectrometer was upgraded together with installation of the W divertor for the purpose of monitoring the W density in the core plasma. We have identified W
, W
and Mo
lines by comparing the spectrum observed by this spectrometer and that calculated by an atomic structure code. Further, from the intensities of these spectral lines, the W density and the Mo density over an electron density were determined to be 
and 
, respectively. The determined W density was in good agreement with that from another spectrometer. In addition, the plasma effective charge evaluated from the continuum intensity observed by the X-ray spectrometer was in agreement with that from a visible spectrometer within a factor of three. From these agreements, it is probable that the determined W and Mo density are valid. In contrast, the determined W density is only 15% compared with that from an soft X-ray array measurement. This inconsistency will be investigated in the near future.
Shiraishi, Junya; Miyato, Naoaki; Matsunaga, Go
no journal, ,
no abstracts in English
Hayashi, Nobuhiko; Jeronimo, G.*; Honda, Mitsuru; Shimizu, Katsuhiro; Hoshino, Kazuo; Ide, Shunsuke; Gerardo, G.*; Sakamoto, Yoshiteru; Suzuki, Takahiro; Urano, Hajime
no journal, ,
no abstracts in English
Honda, Mitsuru; Satake, Shinsuke*; Suzuki, Yasuhiro*; Yoshida, Maiko; Hayashi, Nobuhiko; Kamiya, Kensaku; Matsuyama, Akinobu; Shinohara, Koji; Matsunaga, Go; Nakata, Motoki; et al.
no journal, ,
no abstracts in English
Yamamoto, Tsuyoshi; Hatae, Takaki; Yatsuka, Eiichi; Sugie, Tatsuo; Ogawa, Hiroaki; Takeuchi, Masaki; Kitazawa, Sin-iti; Ota, Kazuya*; Hashimoto, Yasunori*; Nakamura, Kitaru*; et al.
no journal, ,
no abstracts in English
Yoshida, Masafumi; Hanada, Masaya; Kojima, Atsushi; Kashiwagi, Mieko; Umeda, Naotaka; Hiratsuka, Junichi; Akino, Noboru; Endo, Yasuei; Komata, Masao; Mogaki, Kazuhiko; et al.
no journal, ,
One of the key issues for producing high current beams in JT-60SA is a uniform production of the negative ions over the large ion extraction area. Non-uniformity of the negative ion beams in the JT-60 negative ion source was improved by modifying the magnetic filter in the source from the PG filter to a tent-shaped filter. In this work, in order to clarify the influence of the magnetic filter modification on arc efficiency and co-electron current is investigated by means of measurements of H ion, H atoms, co-electron current and beam intensity. As a result, the arc efficiency increase due to increase of H ions and H atoms by reduction of filter field. Although the higher electrons current was extracted than the original PG filter, the co-extracted electron could be suppressed by optimization of the bias voltage. This result contributes to extend the pulse duration time up to 100 s for JT-60SA.
Miyata, Yoshiaki; Suzuki, Takahiro; Ide, Shunsuke; Urano, Hajime; Takechi, Manabu
no journal, ,
It is essential for a stable plasma equilibrium control to reconstruct an accurate plasma boundary in tokamak devices. Cauchy Condition Surface (CCS) method is a numerical approach to calculate the spatial distribution of the magnetic flux outside a hypothetical plasma surface and reconstruct the plasma boundary from the magnetic measurements located outside the plasma. It is found that the optimum number of unknown parameters and shape of the CCS for minimizing errors in the reconstructed plasma shape increase in proportion to the plasma size. It is shown that the accuracy of the plasma shape reconstruction greatly improves by using the optimum number of unknown parameters. The assessment of accuracy of plasma shape reconstruction by CCS method in JT-60SA is reported.
Nakamura, Shigetoshi; Shibama, Yusuke; Masaki, Kei; Sakasai, Akira
no journal, ,
no abstracts in English
Shimizu, Katsuhiro; Hoshino, Kazuo; Kawashima, Hisato; Takizuka, Tomonori*
no journal, ,
no abstracts in English