Overview of materials research and IFMIF-EVEDA under the Broader Approach framework

Nishitani, Takeo; Tanigawa, Hiroyasu; Yamanishi, Toshihiko; Clement Lorenzo, S.*; Baluc, N.*; Hayashi, Kimio; Nakajima, Noriyoshi*; Kimura, Haruyuki; Sugimoto, Masayoshi; Heidinger, R.*; et al.

Fusion Science and Technology, 62(1), p.210 - 218, 2012/07

https://doi.org/10.13182/FST12-A14137

Recent progress in the material related researches and the IFMIF/EVEDA project, which are carried out under the Broader Approach (BA) framework, is reported. In the International Fusion Energy Research Center (IFERC) project of BA, the R&D building was completed March 2010 at the Rokkasho BA site. R&Ds on reduced activation ferritic/ martensitic (RAFM) steels as structural material, SiC/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology relevant to the DEMO operational condition are progressed in Japan and EU. In the IFMIF/EVEDA project, the fabrication of the injector for the IFMIF prototype accelerator was completed at the CEA Saclay, and the first proton beam was obtained in May, 2011. The IFMIF lithium target test loop was completed in March 2011, and a lithium flow of 5 m/s was obtained.

Progress of IFERC project in the Broader Approach activities

Araki, Masanori; Sakamoto, Yoshiteru; Hayashi, Kimio; Nishitani, Takeo; Ouchi, Rei*

Fusion Engineering and Design, 85(10-12), p.2196 - 2202, 2010/12

https://doi.org/10.1016/j.fusengdes.2010.08.029

For contributing to the ITER project and promoting a possible early realization of DEMO, the IFERC project shall perform the activities on (1) DEMO Design and R&D Coordination, (2) Computational Simulation Centre, and (3) ITER Remote Experimentation Centre in the framework under the BA agreement. The DEMO design activity aims at establishing a common basis for DEMO design, including design features of DEMO, a possible common concept of DEMO, a roadmap for DEMO, and so on. Based on the common interest toward DEMO, the DEMO R&D activities have been planned and carried out for the following five areas which are relevant to blanket development: (1) SiC/SiC composites, (2) tritium technology, (3) materials engineering for DEMO blanket, (4) advanced neutron multiplier for DEMO blanket, and (5) advanced tritium breeders for DEMO blanket. In the activity of the Computational Simulation Centre, the objective is to provide and exploit a supercomputer for large scale simulation activities to analyse experimental data on fusion plasmas, prepare scenarios for ITER operation, predict the performance of the ITER facilities and contribute to the DEMO design. At the initial phase, high-level benchmark codes in the fusion research field have been selected through Special Working Group.

New synthesis method of advanced lithium titanate with LiTiO additives for ITER-TBM

Hoshino, Tsuyoshi; Kato, Kenichi*; Natori, Yuri*; Nakamura, Mutsumi*; Sasaki, Kazuya*; Hayashi, Kimio; Terai, Takayuki*; Tatenuma, Katsuyoshi*

Fusion Engineering and Design, 84(2-6), p.956 - 959, 2009/06

https://doi.org/10.1016/j.fusengdes.2008.12.049

LiTiO is one of the most promising candidates among the proposed solid breeder materials for fusion reactors. Addition of H to inert sweep gas has been proposed for enhancing the release of bred tritium from breeder material. However, the mass of LiTiO was found to decrease with time in the hydrogen atmosphere. This mass change indicates that the oxygen content of the sample decreased, suggesting the change from Ti to Ti. In order to control the mass change at the time of high temperature use, the development of lithium titanate which has LiTiO additive is expected to be effective.

Crystal structure of advanced lithium titanate with lithium oxide additives

Hoshino, Tsuyoshi; Sasaki, Kazuya*; Tsuchiya, Kunihiko; Hayashi, Kimio; Suzuki, Akihiro*; Hashimoto, Takuya*; Terai, Takayuki*

Journal of Nuclear Materials, 386-388, p.1098 - 1101, 2009/04

https://doi.org/10.1016/j.jnucmat.2008.12.289

no abstracts in English

Preliminary test for reprocessing technology development of tritium breeders

Hoshino, Tsuyoshi; Tsuchiya, Kunihiko; Hayashi, Kimio; Nakamura, Mutsumi*; Terunuma, Hitoshi*; Tatenuma, Katsuyoshi*

Journal of Nuclear Materials, 386-388, p.1107 - 1110, 2009/04

https://doi.org/10.1016/j.jnucmat.2008.12.297

no abstracts in English

Fusion materials development program in the Broader Approach activities

Nishitani, Takeo; Tanigawa, Hiroyasu; Jitsukawa, Shiro; Nozawa, Takashi; Hayashi, Kimio; Yamanishi, Toshihiko; Tsuchiya, Kunihiko; Mslang, A.*; Baluc, N.*; Pizzuto, A.*; et al.

Journal of Nuclear Materials, 386-388, p.405 - 410, 2009/04

https://doi.org/10.1016/j.jnucmat.2008.12.146

The establishment of the breeding blanket technology is one of the most important engineering issues on the DEMO development. For the DEMO blanket, developments of the structural materials and functional materials such as tritium breeder and neutron multiplier. Which should be used under the savior circumstance such as high neutron fluence, high temperature and strong magnetic field, are urgent issues. In the Broader Approach activities initiated by EU and Japan, developments of reduced activation ferritic martensitic steels as a DEMO blanket structural material, SiC/SiC composites, advanced tritium breeders and neutron multiplier for DEMO blankets, are planed as common interest issues of EU and Japan. This paper describes the overview of the development program.

Investigation and design of the dismantling process of irradiation capsules containing tritium, 2; Detailed design and trial fabrication of capsule dismantling apparatus and investigation of glove box facility

Hayashi, Kimio; Nakagawa, Tetsuya; Onose, Shoji; Ishida, Takuya; Nakamichi, Masaru; Katsuyama, Kozo; Iwamatsu, Shigemi; Hasegawa, Teiji; Kodaka, Hideo; Takatsu, Hideyuki; et al.

JAEA-Technology 2009-007, 168 Pages, 2009/03

JAEA-Technology-2009-007.pdf:31.88MB

In-pile functional tests of breeding blankets have been planned by Japan Atomic Energy Agency (JAEA), using a test blanket module (TBM) which will be loaded in the International Thermonuclear Experimental Reactor (ITER). In preparation for the in-pile functional tests, JAEA has been being performed irradiation experiments of lithium titanate (LiTiO), which is the first candidate of solid breeder materials for the blanket of the demonstration reactor (DEMO) under designing in Japan. The present report describes (1) results of a detailed design and trial fabrication tests of a dismantling apparatus for irradiation capsules which were used in irradiation experiments by the Japan Materials Testing Reactor (JMTR) of JAEA, and (2) results of a preliminary investigation of a glove box facility for post-irradiation examinations (PIEs). In the detailed design of the dismantling apparatus, datailed specifications and the installation methods were examined, based on results of a conceptual design and basic design. In the trial fabrication, cutting tests were curried out by making a mockup of a cutting component. Furthermore, a preliminary investigation of a glove box facility was carried out in order to secure a facility for PIE work after the capsule dismantling, which revealed a technical feasibility.

Investigation and design of the dismantling process of irradiation capsules containing tritium, 1; Conceptual investigation and basic design

Hayashi, Kimio; Nakagawa, Tetsuya; Onose, Shoji; Ishida, Takuya; Kodaka, Hideo; Katsuyama, Kozo; Kitajima, Toshio; Takahashi, Kozo; Tsuchiya, Kunihiko; Nakamichi, Masaru; et al.

JAEA-Technology 2008-010, 68 Pages, 2008/03

JAEA-Technology-2008-010.pdf:11.31MB

In-pile functional tests of breeding blankets for fusion reactors have been planned by Japan Atomic Energy Agency (JAEA), using a test blanket module (TBM) which will be loaded in ITER. The present report describes a conceptual investigation and a basic design of the dismantling process for irradiation capsules which were used in irradiation experiments by the Japan Materials Testing Reactor (JMTR) of JAEA. In the present design, the irradiation capsule is cut by a band saw; the released tritium is recovered safely by a purge-gas system, and is consolidated into a radioactive waste form. Furthermore, adoption of the inner-box enclosing the dismantling apparatus has brought a prospect to be able to utilize an existing hot cell (beta- cell) equipped with usual wall material permeable to tritium, without extensive refurbishing of the cell. Thus, the present study has indicated the feasibility of the present dismantling process for the irradiated JMTR capsules containing tritium.

Research and development of nuclear fusion

Ushigusa, Kenkichi; Seki, Masahiro; Ninomiya, Hiromasa; Norimatsu, Takayoshi*; Kamada, Yutaka; Mori, Masahiro; Okuno, Kiyoshi; Shibanuma, Kiyoshi; Inoue, Takashi; Sakamoto, Keishi; et al.

Genshiryoku Handobukku, p.906 - 1029, 2007/11

no abstracts in English

In-pile tritium permeation through F82H steel with and without a ceramic coating of CrO-SiO Including CrPO

Nakamichi, Masaru; Kulsartov, T. V.*; Hayashi, Kimio; Afanasyev, S. E.*; Shestakov, V. P.*; Chikhray, Y. V.*; Kenzhin, E. A.*; Kolbaenkov, A. N.*

Fusion Engineering and Design, 82(15-24), p.2246 - 2251, 2007/10

https://doi.org/10.1016/j.fusengdes.2007.05.064

no abstracts in English