Japanese contribution to the DEMO-R&D program under the Broader Approach activities

Nishitani, Takeo; Yamanishi, Toshihiko; Tanigawa, Hiroyasu; Nozawa, Takashi; Nakamichi, Masaru; Hoshino, Tsuyoshi; Koyama, Akira*; Kimura, Akihiko*; Hinoki, Tatsuya*; Shikama, Tatsuo*

Fusion Engineering and Design, 86(12), p.2924 - 2927, 2011/12

https://doi.org/10.1016/j.fusengdes.2011.07.004

Several technical R&D activities related to the blanket materials are newly launched as a part of the Broader Approach (BA) activities, which was initiated by the EU and Japan. According to the common interests of these parties for DEMO, R&Ds on reduced activation ferritic/martensitic (RAFM) steels as structural material, SiCf/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out through the BA DEMO R&D program, in order to establish the technical bases on the blanket materials and the tritium technology required for DEMO design. This paper describes overall schedule of those R&D activities and recent progress in Japan carried out by JAEA as the domestic implementing agency on BA, collaborating with Japanese universities and other research institutes.

Validation of welding technology for ITER TF coil structures

Chida, Yutaka; Iguchi, Masahide; Takano, Katsutoshi; Nakajima, Hideo; Osemochi, Koichi*; Niimi, Kenichiro*; Tokai, Daisuke*; Gallix, R.*

Fusion Engineering and Design, 86(12), p.2900 - 2903, 2011/12

https://doi.org/10.1016/j.fusengdes.2011.06.004

TF coil structures, which support large electromagnetic force generated in TF coils under the cryogenic temperature (about 4K), are the mega welding structures composed of coil case and support structures made of high strength and high toughness stainless steel. JAEA started the study on welding trials for heavy thickness materials since 2008 and is planning of full scale mock-up model fabrication for main sub-components (1 set of inboard side and 1set of outboard side) in 2010 in order to investigate the technical issues for manufacturing of TF coil structures. This paper introduces the results on welding trials and status of full scale mock-up model fabrication to confirm the validity of welding technology and manufacturing design before fabricating actual products.

Progress of mock-up trials for ITER TF coil procurement in Japan

Matsui, Kunihiro; Koizumi, Norikiyo; Hemmi, Tsutomu; Takano, Katsutoshi; Nakajima, Hideo; Osemochi, Koichi*; Savary, F.*

Fusion Engineering and Design, 86(6-8), p.1531 - 1536, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.01.053

The magnet system for ITER comprises 18 Toroidal Field (TF) Coils using NbSn cable-in-conduit superconductor, which operate at 4.5 K in supercritical helium. Japan Atomic Energy Agency (JAEA) is responsible for the procurement of 9 TF coils as Japanese Domestic Agency (JADA). Before launching the procurement of these coils, reduced and full-scale trials will be performed to determine and optimize the manufacturing process of a TF coil. During the manufacture of the TF coil, heat-treated superconducting cable-in-conduit conductor, whose length may vary during heat treatment, shall be inserted in the grooves of the radial plate (RP), which is part of the mechanical structure supporting the large electromagnetic forces that are of the order of 800 kN/m. The RP also enhance reliability of the electrical insulation that will be tested up to 19 kV DC and 2.5 kV AC for the winding pack to ground. Very accurate tolerances, of the order of 0.01% on the length of the RP grooves and of the wound conductor, are required to enable the insertion of the conductor. Therefore, the development of suitable manufacturing techniques for the RP and for the winding operation is essential to achieve this requirement. JAEA has contracted companies for fabrication trials of a full-scale RP and winding trials of a one-third scale double pancake to verify feasibility of the required tolerances from an industrial view point. Prior to these trials, JAEA developed a preliminary manufacturing plan and then, industry will carry out small-scale trials to demonstrate applicability of the preliminary manufacturing plan before making the reduced and full-scale trials. The small scale trials will include the cover plate welding with the laser welding, the impregnation using the acryl and metal models, and, the mechanical test and the trail bending of the TF conductor. The results of the small-scale trials and progress on the reduced and full-scale trials are presented in this paper.

Development of Pt/ASDBC catalyst for room temperature recombiner of atmosphere detritiation system

Iwai, Yasunori; Sato, Katsumi; Yamanishi, Toshihiko

Fusion Engineering and Design, 86(9-11), p.2164 - 2167, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.04.028

We have developed the hydrophobic Pt catalysts applicable for tritium oxidation in the presence of saturated water vapor at room temperature. A new type of hydrophobic catalyst, Pt/ASDBC, has been prepared by dipositing platinum on alkyl-styrene diviyl-benzene copolymer (ASDBC). Pt/ASDBC is more hydrophobic than Pt/SDBC that is a promising catalyst for the water detritiation system. The deposited platinum used to prepare Pt/ASDBC catalyst was 1.0 g/L. The value was approximately half of a commercial Japanese Pt/SDBC catalyst. Tritium oxidation tests of the catalysts using 3 GBq/m of tritium were performed in the absence/presence of saturated water vapor at room temperature. Tritium oxidation sufficient for room temperature recombiner was demonstrated using Pt/ASDBC catalyst.

High-efficiency technology for lithium isotope separation using an ionic-liquid impregnated organic membrane

Hoshino, Tsuyoshi; Terai, Takayuki*

Fusion Engineering and Design, 86(9-11), p.2168 - 2171, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.03.090

The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 (Li) in tritium breeding materials. However, natural Li contains only about 7.6 at.% Li. In Japan, new lithium isotope separation technique using ionic-liquid impregnated organic membranes have been developed. The improvement in the durability of the ionic-liquid impregnated organic membrane is one of the main issues for stable, long-term operation of electrodialysis cells while maintaining good performance. Therefore, we developed highly-durable ionic-liquid impregnated organic membrane. Both ends of the ionic-liquid impregnated organic membrane were covered by a nafion 324 overcoat to prevent the outflow of the ionic liquid. The transmission of Lithium aqueous solution after 10 hours under the highly-durable ionic-liquid impregnated organic membrane is almost 13%. So this highly-durable ionic-liquid impregnated organic membrane for long operating of electrodialysis cells has been developed through successful prevention of ion liquid dissolution.

Design of purification loop and traps for the IFMIF/EVEDA Li test loop; Design of cold trap

Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Ida, Mizuho; Yagi, Juro*; Suzuki, Akihiro*; Fukada, Satoshi*; Matsushita, Izuru*; Nakamura, Kazuyuki

Fusion Engineering and Design, 86(9-11), p.2437 - 2441, 2011/10

https://doi.org/10.1016/j.fusengdes.2010.12.021

Engineering Validation and Engineering Design Activities (EVEDA) for The International Fusion Materials Irradiation Facility (IFMIF) were started from July 2007 under an international agreement called ITER Broader Approach. As a major Japanese activity, EVEDA Li test loop (ELTL) to simulate hydraulic and impurity conditions of IFMIF has already designed and is under construction, in which feasibility of hydraulic stability of the liquid Li target, the purification systems of hot traps are major key issues to be validated in this loop. This paper focuses on the purification systems of the ELTL. Design of a cold trap and hot traps are discussed in this paper.

Recent status of fabrication technology development of water cooled ceramic breeder test blanket module in Japan

Hirose, Takanori; Tanigawa, Hisashi; Yoshikawa, Akira; Seki, Yohji; Tsuru, Daigo; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato

Fusion Engineering and Design, 86(9-11), p.2265 - 2268, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.02.024

As one of the most important fabrication technologies of the WCCB TBM, Hot Isostatic Pressing (HIP) joining technology was selected to fabricate the first wall with built-in cooling channel structure made of reduced activation martensitic/ferritic steel, F82H. By using developed HIP technology, a real scale TBM first wall mockup was successfully fabricated. High heat flux test of the fabricated mockup showed the feasibility to with the equivalent conditions of the WCCB TBM operation. The breeder pebble box was successfully fabricated with thin wall cooling pipes and thin plate sleds by Laser welding. With respect to the side walls with built in cooling channels were also fabricated using drilling technology. Assembling of the first wall and side walls is one of the critical fabrication processes of the fabrication of the TBM structure. By using a F82H first wall mockup and side wall mockups, assembling process was demonstrated successfully by Electron Beam welding.

Trial fabrication of beryllides as advanced neutron multiplier

Nakamichi, Masaru; Yonehara, Kazuo; Wakai, Daisuke

Fusion Engineering and Design, 86(9-11), p.2262 - 2264, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.03.009

Design and manufacturing of JT-60SA vacuum vessel

Masaki, Kei; Shibama, Yusuke; Sakurai, Shinji; Katayama, Masahiro*; Sakasai, Akira

Fusion Engineering and Design, 86(9-11), p.1872 - 1876, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.02.008

JT-60SA vacuum vessel (VV) has the outer diameter of 10 m and the height of 6.6 m. The VV is supported by 9 legs. The material is 316L with low cobalt content of 0.05wt%. The VV has a double wall structure composed of inner/outer shells and ribs to ensure high rigidity at operational load and high toroidal one-turn resistance of 16 simultaneously. The double wall thicknesses are 194 mm at inboard and 242 mm at outboard. Inner/outer shells have 18-mm thicknesses. In the double wall, boric-acid water of 50C circulates at plasma operation to reduce nuclear heating of the superconducting coils. At the baking of 200C, nitrogen gas circulates in the double wall. Fundamental welding R&D and a trial manufacturing of the 20 upper half of the VV have been performed to study the manufacturing procedure. After the confirmation of the quality of the mock-up, manufacturing of the actual VV started in December 2009.

Neutronic analysis of the ITER poloidal polarimeter

Ishikawa, Masao; Kawano, Yasunori; Imazawa, Ryota; Sato, Satoshi; Vayakis, G.*; Bertalot, L.*; Yatsuka, Eiichi; Hatae, Takaki; Kondoh, Takashi; Kusama, Yoshinori

Fusion Engineering and Design, 86(6-8), p.1286 - 1289, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.01.046

The nuclear heating rates of the optical mirrors of the poloidal polarimeter installed in the equatorial port plug of ITER are calculated. Since the system cannot have a sufficiently labyrinthine structure and the second mirrors are located nearly as close to the plasma as the first mirrors due to limited space, the nuclear heating rate of the second mirrors is as high as that of the first mirrors. However, it is possible to reduce the nuclear heating rates of the mirrors if the blanket shield module provides a sufficient degree of neutron shielding.