Suppression of radiation-induced point defects by rhenium and osmium interstitials in tungsten

Suzudo, Tomoaki; Hasegawa, Akira*

Scientific Reports (Internet), 6, p.36738_1 - 36738_6, 2016/11

https://doi.org/10.1038/srep36738

Modeling of the evolution of radiation-induced defects is important for finding radiation-resistant materials, which would be greatly appreciated in nuclear applications. We apply the first principles method combined with kinetic Monte Carlo to indicate a mechanism to mitigate the effect of radiation by adding particular solute elements that change the migration dimension of interstitials in W crystals. The resultant mechanism is applicable to any body-centered-cubic (BCC) metals whose SIAs have one-dimensional (1D) motion and is expected to provide a general guideline for computational design of radiation-resistant alloys in the field of nuclear applications.

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.

Depth profile measurements of hydrogen isotopes near the surface of the TFTR plasma facing component using nuclear reaction analysis

Kubota, Naoyoshi; Ochiai, Kentaro; Kutsukake, Chuzo; Hayashi, Takao; Shu, Wataru; Nishi, Masataka; Nishitani, Takeo

Purazuma, Kaku Yugo Gakkai-Shi, 81(4), p.296 - 301, 2005/04

http://ci.nii.ac.jp/naid/110006281931

no abstracts in English

Trailing the hydrogen isotopes in the nuclear fusion device

Masaki, Kei

Nihon Genshiryoku Gakkai-Shi, 46(12), p.834 - 837, 2004/12

http://ci.nii.ac.jp/naid/10013704958

no abstracts in English

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.

Development of the level difference measuring method by the coarseness meter

Yagisawa, Hiroshi; Arai, Takashi; Goto, Yoshitaka*; Kaminaga, Atsushi; Miya, Naoyuki

KEK Proceedings 2003-16 (CD-ROM), 4 Pages, 2004/02

no abstracts in English

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

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

https://doi.org/10.1238/Physica.Topical.091a00113

no abstracts in English