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.

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.

Repetitive gyrotron operation for ITER

Kajiwara, Ken; Oda, Yasuhisa; Kasugai, Atsushi; Takahashi, Koji; Sakamoto, Keishi; Darbos, C.*; Henderson, M. A.*

Fusion Engineering and Design, 86(6-8), p.955 - 958, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.01.084

Thermo-structural analysis of integrated back plate in IFMIF/EVEDA liquid lithium target

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

Fusion Engineering and Design, 86(9-11), p.2482 - 2486, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.01.038

The Engineering Validation and Engineering Design Activities (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) are in progress under the Broader Approach (BA) Agreement. As a part of this engineering design, we carried out thermo-structural analysis of the back plate in the IFMIF target. In this analysis, the target assembly of the integrated back plate option was modeled with the nuclear heating to simulate the IFMIF usual operation. The calculation parameters were thermal boundary conditions of a mechanical joint between the target assembly and the beam duct. The calculation results showed the influence of parameters on thermal stress was small. The maximum von Mises stresses occurred at the back plate center and those values, 204 - 218 MPa were lower than half of the yield strength of F82H (455 MPa). The maximum thermal deformations occurred at the same place and those values, about 0.3 mm will be important input parameter for the Li flow stability analysis.

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.

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

Wave period of free-surface waves on high-speed liquid lithium jet for IFMIF target

Kanemura, Takuji; Sugiura, Hirokazu*; Yamaoka, Nobuo*; Suzuki, Sachiko*; Kondo, Hiroo; Ida, Mizuho; Matsushita, Izuru*; Horiike, Hiroshi*

Fusion Engineering and Design, 86(9-11), p.2462 - 2465, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.01.030

Wave period of free-surface waves on a high-speed liquid lithium (Li) jet is very important wave characteristics to investigate for validation of a Li target of the International Fusion Materials Irradiation Facility (IFMIF). In this paper, we report characteristics of wave period measured by a contact-type liquid level sensor. The experiments were conducted at a Li loop in Osaka University. In this loop, a plane Li jet simulating the IFMIF Li target can be controlled at the velocities of up to 15 m/s. Probability density distribution of the measured wave periods was nearly equal to the log-normal distribution. The fact that the wave period distribution is nearly equal to the log-normal distribution has been already identified in the ocean waves which are known for its random property. From present and previous our experimental results, it was concluded that random wave property developed for the ocean waves can apply to the free-surface waves on the Li jet.