Long-term properties of reduced activation ferritic/martensitic steels for fusion reactor blanket system

Shiba, Kiyoyuki; Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro

Fusion Engineering and Design, 86(12), p.2895 - 2899, 2011/12

https://doi.org/10.1016/j.fusengdes.2011.06.005

Aging properties of reduced activation ferritic/martensitic steel F82H was researched at temperature ranging from 400C to 650C up to 100,000 hr. Microstructure, tensile, and Charpy properties were carried out. Laves was found at temperatures between 550 and 650C and MC carbides were found at the temperatures between 500 and 600C over 10,000 hr. These precipitates caused degradation in toughness, especially at temperatures ranging from 550C to 650C. Tensile properties do not have serious aging effect, except for 650C, which caused large softening even after 10.000 hr. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in DBTT caused by the large Laves phase precipitation at grain boundary was observed in the 650C aging. Laves precipitates at grain boundary also degrades the USE of the aged materials. These aging test results provide F82H can be used up to 30,000 hr at 550C.

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.

Simplification of blanket system for SlimCS fusion DEMO reactor

Someya, Yoji; Takase, Haruhiko; Uto, Hiroyasu; Tobita, Kenji; Liu, C.; Asakura, Nobuyuki

Fusion Engineering and Design, 86(9-11), p.2269 - 2272, 2011/11

https://doi.org/10.1016/j.fusengdes.2011.01.141

Conceptual design of a tritium-breeding blanket for SlimCS has been studied. The blanket structure with neutron multiplier Be-plate was designed to be as thin as possible with keeping high Tritium Breeding Ratio (TBR). However, a structure of the blanket is complexity and the manufacture of the blanket is difficult from the viewpoint of engineering. Therefore, simplification of blanket structure is necessary for SlimCS. In this paper, we propose a simple blanket structure without decreasing the net TBR below 1.05. The proposed blanket structure is mixed LiSiO pebbles or LiO pebbles for the tritium breeding and BeTi pebbles for the neutron multiplication and these pebbles are filled in the blanket. ANIHEAT code with the nuclear data library FENDL-2.0 was used for the calculations of the neutronics and thermal analyses. As a result, it is shown that LiO pebbles blanket mixed with BeTi pebbles is the most effective and the TBR is greater than 1.05.

First qualification of ITER toroidal field coil conductor jacketing

Hamada, Kazuya; Takahashi, Yoshikazu; Isono, Takaaki; Nunoya, Yoshihiko; Matsui, Kunihiro; Kawano, Katsumi; Oshikiri, Masayuki; Tsutsumi, Fumiaki; Koizumi, Norikiyo; Nakajima, Hideo; et al.

Fusion Engineering and Design, 86(6-8), p.1506 - 1510, 2011/10

https://doi.org/10.1016/j.fusengdes.2010.12.054

Japan Atomic Energy Agency has a responsibility for procurement of the ITER toroidal field coil conductors as Japanese Domestic Agency (JADA) of the ITER project. The TF conductor is a circular shaped cable-in-conduit conductor, which is composed of cable and stainless steel conduit (jacket). The outer diameter and wall thickness of jacket are 43.7mm and 2mm, respectively. The cable consists of 900 NbSn superconducting strands and 522 Cu strands. The length of TF conductor is 780m in maximum. Preparation of conductor fabrication was completed in December 2009. And then, to demonstrate a conductor manufacturing procedure, JADA fabricated 780m-long Cu dummy conductor as a process qualification. Finally, the 780m-long Cu dummy conductor has been successfully completed, ahead of other domestic agencies that are in charge of TF conductor procurement. Since all of manufacturing processes have been qualified, JADA started to fabricate superconducting conductors for TF coils.

Trial fabrication tests of advanced tritium breeder pebbles using sol-gel method

Hoshino, Tsuyoshi; Oikawa, Fumiaki

Fusion Engineering and Design, 86(9-11), p.2172 - 2175, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.01.065

Lithium titanate (LiTiO) is one of the most promising candidates among tritium breeding materials because of its good tritium release characteristics. However, the mass of LiTiO decreased with time in a hydrogen atmosphere by Li evaporation and with Li burn up. In order to prevent the mass decrease at high temperatures, LiTiO with added Li have been developed as one of advanced tritium breeders. We have been promoting the development of fabrication technique of LiTiO pebbles by the sol-gel method. The fabrication techniques of advanced tritium breeder pebbles have not been established for large quantities. Therefore, trial fabrication tests of advanced breeder pebbles were carried out using previous sol-gel method. The diameter of the pebbles is 1.18 mm, and the sphericity is 1.04. It is expected that an advanced tritium breeder with added Li will be stable under operating conditions, namely in a neutron environment at a high temperatures. Thus, these results show that the pebble fabrication using the sol-gel method is a promising production technique for mass production of the advanced tritium breeder pebbles.

Present status of Japanese tasks for lithium target facility under IFMIF/EVEDA

Nakamura, Kazuyuki; Furukawa, Tomohiro; Hirakawa, Yasushi; Kanemura, Takuji; Kondo, Hiroo; Ida, Mizuho; Niitsuma, Shigeto; Otaka, Masahiko; Watanabe, Kazuyoshi; Horiike, Hiroshi*; et al.

Fusion Engineering and Design, 86(9-11), p.2491 - 2494, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.04.033

In IFMIF/EVEDA, tasks for lithium target system are shared to 5 validation tasks (LF1-5) and a design task (LF6). The purpose of LF1 task is to construct and operate the EVEDA lithium test loop, and JAEA has a main responsibility to the performance of the Li test loop. LF2 is a task for the diagnostics of the Li test loop and IFMIF design. Basic research for the diagnostics equipment has been completed, and the construction for the Li test loop will be finished before March in 2011. LF4 is a task for the purification systems with nitrogen and hydrogen. Basic research for the purification equipment has been completed, and the construction of the nitrogen system for the Li test loop will be finished before March in 2011. LF5 is a task for the remote handling system with the target assembly. JAEA has an idea to use the laser beam for cutting and welding of the lip part of the flanges. LF6 is a task for the design of the IFMIF based on the validation experiments of LF1-5.

Safety concept of the IFMIF/EVEDA lithium test loop

Furukawa, Tomohiro; Kondo, Hiroo; Hirakawa, Yasushi; Kato, Shoichi; Matsushita, Izuru*; Ida, Mizuho; Nakamura, Kazuyuki

Fusion Engineering and Design, 86(9-11), p.2433 - 2436, 2011/10

https://doi.org/10.1016/j.fusengdes.2010.12.037

In order to obtain the engineering data on the lithium target system, which is the neutron source of the International Fusion Material Irradiation Facility (IFMIF), the design and fabrication of the IFMIF/EVEDA Lithium Test Loop are being carried out under the Engineering Validation and Engineering Design Activity (EVEDA). The loop will hold 2.5 tons of lithium. Since lithium is specified by Japanese law as a dangerous substance, countermeasures which assumed a lithium leak incident and various abnormal issues are indispensable. This paper describes about the safety principles and measures for lithium leaks of the IFMIF/EVEDA lithium test loop decided under the detailed design process.

Design study of an AC power supply system in JT-60SA

Shimada, Katsuhiro; Baulaigue, O.*; Cara, P.*; Coletti, A.*; Coletti, R.*; Matsukawa, Makoto; Terakado, Tsunehisa; Yamauchi, Kunihito

Fusion Engineering and Design, 86(6-8), p.1427 - 1431, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.02.097

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.

Hydraulic analysis on effects of back-plate deformation upon stability of high-speed free-surface lithium flow for IFMIF target design

Ida, Mizuho; Kondo, Hiroo; Nakamura, Kazuyuki; Wakai, Eiichi

Fusion Engineering and Design, 86(9-11), p.2478 - 2481, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.02.093

For engineering design of liquid lithium (Li) target of the International Fusion Materials Irradiation Facility (IFMIF), two-dimensional hydraulic analysis was performed using code FLUENT. The analysis clarified effects of smooth deformation at the center of the back-plate part of the flow channel upon velocity and pressure in the high-speed free-surface Li flow. Convex deformations caused decrease in jet thickness and thus increase in velocity and decrease in pressure. Convex deformations caused vice versa. Acceptable limit of the back-plate deformation to avoid Li boiling in IFMIF target was 1.0 mm for the convex deformation and 3.0 mm for the concave deformation. Spatial range of the effects of the deformation was localized around deformation as far as no boiling.