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Takeishi, Toshiharu*; Katayama, Kazunari*; Nishikawa, Masabumi*; Masaki, Kei; Miya, Naoyuki
Journal of Nuclear Materials, 349(3), p.327 - 338, 2006/03
Times Cited Count：6 Percentile：40.42(Materials Science, Multidisciplinary)no abstracts in English
Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato; Ibbott, C.*; Tivey, R.*
Fusion Science and Technology, 46(4), p.530 - 540, 2004/12
Times Cited Count：14 Percentile：65.14(Nuclear Science & Technology)The first fabrication and heating test of a large-scale CFC monoblock divertor mock-up using annular flow concept have been performed to demonstrate its manufacturability and thermo-mechanical performance. Prior to the fabrication of the mock-up, brazed joint tests between the CFC monoblock and the CuCrZr tube have been carried out to find the suitable heat treatment mitigating loss of the high mechanical strength of the CuCrZr material. Basic mechanical examination on CuCrZr undergoing the brazing heat treatment and FEM analyses are also performed to support the design of the mock-up. High heat flux tests on the large-scale divertor mock-up have been performed in an ion beam facility. The mock-up has successfully withstood more than 1,000 thermal cycles of 
for 15 s and 3,000 cycles more than 
for 15 s, which simulates the heat load condition of the ITER divertor. No degradation of the thermal performance of the mock-up has been observed throughout the thermal cycle test.
Suzuki, Satoshi
Koon Gakkai-Shi, 30(5), p.243 - 247, 2004/09
Selection and the development of plasma facing materials for fusion devices, mainly ITER, are presented. For the divertor, CFC (Carbon fiber reinforced carbon composite) materials are utilized as plasma facing materials in the lower part of vertical targets in ITER. Since the design maximum heat flux to the vertical targets is 20 MW/m
, CFC materials, which have higher thermal conductivity than pure copper, are preferable from a heat removal point of view. On the other hand, a plasma facing material of a dome and a liner is tungsten because tungsten has low sputtering yield and has relatively high thermal conductivity among metals. First wall covers 80% of the plasma facing area of ITER. The plasma facing material of the first wall should have good compatibility with plasma. Therefore, beryllium is utilized as a plasma facing material from the low contamination and the minimization of the oxygen impurity to the plasma points of view.
Goto, Yoshitaka*
Kaku Yugoro, (11), p.34 - 37, 2004/03
Recent results of erosion and re-deposition studies are reported for graphite tiles from W-shaped divertor region of JT-60U. The tiles were operated in Jun. 1997 - Oct. 1998 periods in which more than 3000 D-D discharge experiments were made with all-carbon walls with 2-times boronizations and inner-private flux pumping. Erosion depth was estimated by using dial gauge, while deposition was measured with SEM. On the outer divertor target, erosion was found dominant, while on the inner target, re-deposition was found to be dominant. No continuous deposition layers on the dome top. Observed in/out asymmetry is attributable to in/out asymmetry of plasma particle conditions in front of the divertor plates. Columnar structures in the redeposition layers corresponded to the lower heat-flux zone, while lamellar structures were found in the overlayers in the higher heat-flux zone near the separatrix strike point.
Nakamura, Hirofumi; Nishi, Masataka
JAERI-Research 2003-024, 24 Pages, 2003/11
JAERI-Research-2003-024.pdf:1.12MB
Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium transport properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authors' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evaluation was also carried out for comparison (previous data). The permeation analysis was carried out individually by classifying into the armor region (Carbon Fiber Composites and tungsten) and the slit region without armor (3% of armor surface area) assuming the incident flux and temperature for each region. As the results of the permeation analysis, estimated permeation amount with the authors' data was one order less than that with the previous data at the end of lifetime of the divertor due to authors' small diffusion coefficient of tritium in tungsten. It also indicated the possibility that permeation through the slit region of the armor tiles could dominate total permeation through the vertical target, since tritium permeation amount through tungsten armor with the authors' data was estimated to be reduced drastically smaller than that with the previous evaluation data. The result of a little tritium permeation amount through the vertical target with the authors' data ensured the conservatism of the current evaluation of tritium concentration in the primary cooling water in ITER divertor, as it indicated the possibility of direct drainage of the divertor primary cooling water.
Masaki, Kei; Yagyu, Junichi; Arai, Takashi; Kaminaga, Atsushi; Kodama, Kozo; Miya, Naoyuki; Ando, Toshiro; Hiratsuka, Hajime; Saido, Masahiro
Fusion Science and Technology (JT-60 Special Issue), 42(2-3), p.386 - 395, 2002/09
Times Cited Count：9 Percentile：50.57(Nuclear Science & Technology)The wall conditioning of JT-60U consists of the 300
C baking, He-TDC, He-GDC, tokamak discharge cleaning and boronization. Using these methods, total pressure of the vacuum vessel reached finally 10
10
Pa. The oxygen impurity was decreased to 0.5%. The experience with the carbon-based first wall showed that taper shaping is effective to reduce the local heat concentration to the tile edges. The observed poloidal asymmetric deposition of carbon on the divertor region implies that the carbon impurity produced in the outer divertor contributes to the deposition on the inner divertor. In 1992 and 1993, the B
C converted CFC tiles were installed in the outer divertor to reduce chemical sputtering of CFC tiles and oxygen impurity. The reduction was successfully demonstrated with the BC converted CFC tiles. In order to understand the tritium behavior in JT-60U, tritium in the first wall and the exhaust gas were measured. The estimated tritium inventory in the first wall was 50% of the generated tritium.
Sato, Kazuyoshi; Ezato, Koichiro; Taniguchi, Masaki; Suzuki, Satoshi; Akiba, Masato
Journal of Plasma and Fusion Research SERIES, Vol.5, p.556 - 560, 2002/00
no abstracts in English
Masaki, Kei; Akiba, Masato
Purazuma, Kaku Yugo Gakkai-Shi, 77(9), p.884 - 893, 2001/09
no abstracts in English
Sato, Kazuyoshi; Ishitsuka, Etsuo; Uchida, Munenori*; Kawamura, Hiroshi; Ezato, Koichiro; Taniguchi, Masaki; Akiba, Masato
Physica Scripta, T91, p.113 - 116, 2001/07
Times Cited Count：1 Percentile：11.82(Physics, Multidisciplinary)no abstracts in English
Ezato, Koichiro; Taniguchi, Masaki; Sato, Kazuyoshi; Araki, Masanori; Akiba, Masato
Physica Scripta, T91, p.110 - 112, 2001/07
Times Cited Count：2 Percentile：20.45(Physics, Multidisciplinary)no abstracts in English
Tadokoro, Takahiro*; Isobe, Kanetsugu; Ohira, Shigeru; Shu, Wataru; Nishi, Masataka
Journal of Nuclear Materials, 283-287(Part2), p.1048 - 1052, 2000/12
Times Cited Count：2 Percentile：19.11(Materials Science, Multidisciplinary)no abstracts in English
Sato, Kazuyoshi; Ishitsuka, Etsuo; Uda, Minoru*; Kawamura, Hiroshi; Suzuki, Satoshi; Taniguchi, Masaki; Ezato, Koichiro; Akiba, Masato
Journal of Nuclear Materials, 283-287(2), p.1157 - 1160, 2000/12
Times Cited Count：6 Percentile：41.99(Materials Science, Multidisciplinary)no abstracts in English
C-carbon fiber composite irradiated with low energy deuterium ionsJimbo, Ryutaro*; Nakamura, Kazuyuki; Bandourko, V.*; Dairaku, Masayuki; Okumura, Yoshikazu; Akiba, Masato
Journal of Nuclear Materials, 266-269, p.1103 - 1107, 1999/00
Times Cited Count：5 Percentile：40.11(Materials Science, Multidisciplinary)no abstracts in English
Yamada, Reiji; Fujii, Kimio
Mater. Sci. Forum, 308-311, p.902 - 907, 1999/00
no abstracts in English
Uda, Minoru*; Ishitsuka, Etsuo; Sato, Kazuyoshi; Akiba, Masato; ; Kawamura, Hiroshi
Phys. Scr., T81, p.98 - 100, 1999/00
Times Cited Count：2 Percentile：26.00(Physics, Multidisciplinary)no abstracts in English
Masaki, Kei; ; ; Morimoto, Masaaki*; ; Hosogane, Nobuyuki; Sakurai, Shinji; Saido, Masahiro
Purazuma, Kaku Yugo Gakkai-Shi, 74(9), p.1048 - 1053, 1998/09
no abstracts in English
Masaki, Kei; ; Morimoto, Masaaki*; ; ; Hosogane, Nobuyuki; Saido, Masahiro
Fusion Technology 1998, p.67 - 70, 1998/00
no abstracts in English
Sato, Kazuyoshi; ; ; Nakamura, Kazuyuki; Araki, Masanori; Akiba, Masato
Fusion Technology 1998, p.109 - 112, 1998/00
no abstracts in English
Uda, Minoru*; Ishitsuka, Etsuo; Sato, Kazuyoshi; Akiba, Masato; ; ; Kawamura, Hiroshi
Fusion Technology 1998, 1, p.161 - 164, 1998/00
no abstracts in English
Nakamura, Kazuyuki; ; Dairaku, Masayuki; ; Okumura, Yoshikazu; ; Jimbo, Ryutaro*; Bandourko, V.*; Akiba, Masato
Journal of Nuclear Materials, 258-263, p.828 - 832, 1998/00
Times Cited Count：4 Percentile：38.45(Materials Science, Multidisciplinary)no abstracts in English
