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Journal Articles

Tritium release from bulk of carbon-based tiles used in JT-60U

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:43.58(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Development of ITER divertor vertical target with annular flow concept, 2; Development of brazing technique for CFC/CuCrZr joint and heating test of large-scale mock-up

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:68.23(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 $$rm 20 MW/m^2$$ for 15 s and 3,000 cycles more than $$rm 10 MW/m^2$$ 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.

Journal Articles

First wall and divertor materials as plasma facing components

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$$^{2}$$, 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.

Journal Articles

Erosion and re-deposition in JT-60U

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.

JAEA Reports

Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties

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.

Journal Articles

Wall conditioning and experience of the carbon-based first wall in JT-60U

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:8 Percentile:49.76(Nuclear Science & Technology)

The wall conditioning of JT-60U consists of the 300$$^{circ}$$C baking, He-TDC, He-GDC, tokamak discharge cleaning and boronization. Using these methods, total pressure of the vacuum vessel reached finally 10$$^{-6}$$ $$sim$$ 10$$^{-5}$$ Pa. The oxygen impurity was decreased to $$sim$$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$$_{4}$$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 B$$_{4}$$C 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 $$sim$$50% of the generated tritium.

Journal Articles

Development of ITER divertor plate with annular swirl tube and tungsten rods

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

Journal Articles

Erosion and redeposition at divertor plate

Masaki, Kei; Akiba, Masato

Purazuma, Kaku Yugo Gakkai-Shi, 77(9), p.884 - 893, 2001/09

no abstracts in English

Journal Articles

Thermal cycle experiments of neutron-irradiated CFC/Cu mock-ups

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:12.24(Physics, Multidisciplinary)

no abstracts in English

Journal Articles

Ultrasonic non-destractive testing on CFC monoblock divertor mock-up

Ezato, Koichiro; Taniguchi, Masaki; Sato, Kazuyoshi; Araki, Masanori; Akiba, Masato

Physica Scripta, T91, p.110 - 112, 2001/07

 Times Cited Count:2 Percentile:21.24(Physics, Multidisciplinary)

no abstracts in English

Journal Articles

Depth profile of tritium in plasma exposed CX-2002U

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:20.05(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Erosion characteristics of neutron-irradiated carbon-based materials under simulated disruption heat loads

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:44.06(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Temperature and angular dependences of sputtering yield of B$$_{4}$$C-carbon fiber composite irradiated with low energy deuterium ions

Jimbo, 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:41.94(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Oxidation and thermal shock resistance of SiC compositionally graded carbon fiber-reinforced carbon composite materials

Yamada, Reiji; Fujii, Kimio

Mater. Sci. Forum, 308-311, p.902 - 907, 1999/00

no abstracts in English

Journal Articles

Thermal shock test of neutron irradiated carbon fiber reinforced carbon composites with OHBIS

Uda, Minoru*; Ishitsuka, Etsuo; Sato, Kazuyoshi; Akiba, Masato; *; Kawamura, Hiroshi

Phys. Scr., T81, p.98 - 100, 1999/00

 Times Cited Count:2 Percentile:26.58(Physics, Multidisciplinary)

no abstracts in English

Journal Articles

Inspection of JT-60 W-shaped divertor after the initial operation

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

Journal Articles

Development of plasma facing components for fusion experimental reactors in JAERI

Sato, Kazuyoshi; ; *; Nakamura, Kazuyuki; Araki, Masanori; Akiba, Masato

Fusion Technology 1998, p.109 - 112, 1998/00

no abstracts in English

Journal Articles

Disruption and erosion on plasma facing materials with Oarai hot-cell electron beam irradiating system (OHBIS)

Uda, Minoru*; Ishitsuka, Etsuo; Sato, Kazuyoshi; Akiba, Masato; *; *; Kawamura, Hiroshi

Fusion Technology 1998, 1, p.161 - 164, 1998/00

no abstracts in English

Journal Articles

The First inspection of JT-60U W-shaped divertor after high power operation

Masaki, Kei; ; Morimoto, Masaaki*; ; *; Hosogane, Nobuyuki; Saido, Masahiro

Fusion Technology 1998, p.67 - 70, 1998/00

no abstracts in English

Journal Articles

Disruption and sputtering erosions on SiC doped CFC

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:39.76(Materials Science, Multidisciplinary)

no abstracts in English

41 (Records 1-20 displayed on this page)