New conceptual design of a test module assembly for tritium permeation experiment

Ohira, Shigeru; Luo, G.; Nakamura, Hirofumi; Shu, Wataru; Kitamura, Kazunori*; Nishi, Masataka

Fusion Science and Technology, 48(1), p.621 - 624, 2005/07

https://doi.org/10.13182/FST05-A1002

A new conceptual design of a tritium permeation test module assembly was developed for demonstration tests using a strong plasma source constructed in TPL for simulation of practical tritium permeation in the real plasma facing components and validation of the models and codes for evaluation of tritium permeation. The target module, to be irradiated by the high flux plasma beam, consists of a multi-layer structure of a plasma facing material plate and a copper substrate with pressurized coolant cavity. Tests using a preliminary model of the target module has been performed to verify thermal and mechanical behavior of the bonded structure and to assess its structural integrity focused on the bonded interface under the cyclic heat loads. After the heat load testing, no visible defect and crack was observed around the bonded interfaces with magnifying glass, and its structural integrity was verified. Also the first tests using tritium plasma at TSTA/LANL with the preliminary model was performed and process to measured tritium permeated was established.

Integrity of the first wall in fusion reactors

Kurihara, Ryoichi

JAERI-Tech 2004-052, 39 Pages, 2004/07

JAERI-Tech-2004-052.pdf:2.1MB

The problems in the thermal structural design of the plasma facing component such as the blanket first wall and the divertor plate which receives very high heat flux were examined in the design of the fusion power reactors. Compact high fusion power reactor must give high heat flux and high-speed neutron flux from the plasma to the first wall and the divertor plate. In this environmental situation, the micro cracks should be generated in material of the first wall. Structural integrity of the first wall would be very low during the operation of the reactor, if those micro-cracks grow in a crack having significant size by the fatigue or the creep. The crack penetration in the first wall can be a factor which threatens the safety of the fusion power reactor. This paper summarizes the problems on the structural integrity in the first wall made of the SiC/SiC composite material or the ferritic steel.

Advanced fusion technologies developed for JT-60 superconducting Tokamak

Sakasai, Akira; Ishida, Shinichi; Matsukawa, Makoto; Akino, Noboru; Ando, Toshinari*; Arai, Takashi; Ezato, Koichiro; Hamada, Kazuya; Ichige, Hisashi; Isono, Takaaki; et al.

Nuclear Fusion, 44(2), p.329 - 334, 2004/02

https://doi.org/10.1088/0029-5515/44/2/015

no abstracts in English

Mechanical properties of HIP bonded W and Cu-alloys joint for plasma facing components

Saito, Shigeru; Fukaya, Kiyoshi*; Ishiyama, Shintaro; Sato, Kazuyoshi

Journal of Nuclear Materials, 307-311(2), p.1542 - 1546, 2002/12

https://doi.org/10.1016/S0022-3115(02)01169-8

no abstracts in English

Tritium decontamination of TFTR D-T plasma facing components using ultra violet laser

Shu, Wataru; Kawakubo, Yukio*; Ohira, Shigeru; Oya, Yasuhisa; Hayashi, Takumi; Nakamura, Hirofumi; Iwai, Yasunori; Nishi, Masataka; Gentile, C. A.*; Skinner, C. H.*; et al.

Fusion Science and Technology, 41(3), p.690 - 694, 2002/05

no abstracts in English

Crack propagation tests of HIPed DSCu/SS joints for plasma facing components

Hatano, Toshihisa; Goto, Masahiro*; Yamada, Tetsuji*; Nomura, Yuichiro*; Saito, Masakatsu*

Fusion Engineering and Design, 49-50, p.207 - 212, 2000/11

https://doi.org/10.1016/S0920-3796(00)00443-9

no abstracts in English

Development of bonding techniques between tungsten and copper alloy for plasma facing components by HIP method, 1; Bonding between tungsten and Oxygen Free Copper

Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Eto, Motokuni; Akiba, Masato

JAERI-Research 99-049, 36 Pages, 1999/08

JAERI-Research-99-049.pdf:3.81MB

no abstracts in English

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

Development of BC-carbon fiber composite ceramics as plasma facing materials in nuclear fusion reactor, part 3; Heat resistance evaluation by electron beam irradiation and by in situ plasma discharge in JT-60

*; Saido, Masahiro; Nakamura, Kazuyuki; Akiba, Masato; ; Dairaku, Masayuki; *; *; *; Goto, Yoshitaka*

Nihon Seramikkusu Kyokai Gakujutsu Rombunshi, 105(1228), p.1091 - 1098, 1997/12

https://doi.org/10.2109/jcersj.105.1091

no abstracts in English

Reactivity test between beryllium and dispersion strengthened copper

*; Kawamura, Hiroshi; R.R.Solomon*

Fusion Technology 1996, 0, p.407 - 410, 1997/00

no abstracts in English