Deuterium behaviour at the interface of oxidized metal under high temperature heavy water

Nakamura, Hirofumi; Hatano, Yuji*; Yamanishi, Toshihiko

Fusion Engineering and Design, 87(5-6), p.916 - 920, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.044

Deuterium behavior in the metal exposed to hot heavy water has been investigated in order to understand the oxidation driven tritium permeation in the fusion reactor. Disks of SS304, F82H and Ni and gold plated SS304 and F82H were oxidized in an autoclave at 573K. After the oxidation, soaked deuterium in the specimen was measured by the thermal desorption method and elemental depth distribution in the specimen was measured by a glow discharge optical elemental spectroscopy method. Obtained results were followings, (1) The oxide thickness has grown with the soaking time, and solved deuterium amount also increases with oxidation time for all materials. (2) Deuterium exists at the interface of the oxide and metal for all materials. (3) Deuterium in the gold plated samples were less than that in the bare SS304 about 1/5. (4) Deuterium in nickel was less than that in the SS304 by one orders magnitude and oxide thickness was also thinner than SS304. Those results indicate that deuterium solution into the material would be initiated by the deuterium gas production at the oxidation process of metal. Gold plating as the oxidation protection could be effective to prevent deuterium solution into the metal.

Development of the water cooled ceramic breeder test blanket module in Japan

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; et al.

Fusion Engineering and Design, 87(7-8), p.1363 - 1369, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.03.007

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. Fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

Overview of R&D activities on tritium processing and handling technology in JAEA

Yamanishi, Toshihiko; Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Oyaizu, Makoto; Yamada, Masayuki; Suzuki, Takumi; Hayashi, Takumi

Fusion Engineering and Design, 87(5-6), p.890 - 895, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.043

In JAEA, the tritium processing and handling technologies have been studied at TPL. The main basic R&D activities in this field are: the tritium processing technology for the blanket recovery system; the tritium behavior in a confinement; and detritiation and decontamination. The R&D for tritium processing and handling technologies to a demonstration reactor (DEMO) are also planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities. The ceramic electrolysis cell has been studied as a tritium processing method for the blanket system. The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column.

Evaluation of -ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

Takahashi, Hiroki; Maebara, Sunao; Sakaki, Hironao; Hirabayashi, Keiichi*; Hidaka, Kosuke*; Shigyo, Nobuhiro*; Watanabe, Yukinobu*; Sagara, Kenshi*

Fusion Engineering and Design, 87(7-8), p.1235 - 1238, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.109

The Engineering Validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of -rays and He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated.

Evaluation of local deformation behavior accompanying fatigue damage in F82H welded joint specimens by using digital image correlation

Nakata, Toshiya; Tanigawa, Hiroyasu

Fusion Engineering and Design, 87(5-6), p.589 - 593, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.01.048

In tensile and fatigue testing, the deformation behavior of local domains of tungsten inert gas (TIG) and electron beam (EB) welded joint specimens of F82H steel was evaluated by using digital image correlation. For the tensile test specimens, tensile strength declined in the TIG welded joint material and ductility declined in both the EB and TIG welded joint materials. Because axial strain increased in the tempered heat-affected zone (HAZ) and led to fracture of the TIG welded joint material, strength was considered to deteriorate. In fatigue testing, the number of cycles to fracture for the welded joint material decreased to less than 40-50% that for the base metal. For both fracture metals, shear strain exhibited the largest value approximately between the fine-grained HAZ and the tempered HAZ, leading to fracture. Cavities and macrocracks formed in the fine-grained HAZ and tempered HAZ in the fracture metal cross section, and geometrical damage possibly resulted in deterioration of fatigue lifetime.

Nuclear engineering of diagnostic port plugs on ITER

Pitcher, C. S.*; Barnsley, R.*; Feder, R.*; Hu, Q.*; Loesser, G. D.*; Lyublin, B.*; Padasalagi, S.*; Pak, S.*; Reichle, R.*; Sato, Kazuyoshi; et al.

Fusion Engineering and Design, 87(5-6), p.667 - 674, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.01.043

Characteristics of surface oscillation on high speed liquid Li jet

Suzuki, Sachiko*; Hoashi, Eiji*; Kanemura, Takuji; Kondo, Hiroo; Yamaoka, Nobuo*; Horiike, Hiroshi*

Fusion Engineering and Design, 87(7-8), p.1434 - 1438, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.03.027

In the international fusion materials irradiation facility (IFMIF), the high speed liquid metal lithium (Li) wall jet will be used as a target irradiated by two deuteron beams to generate intense neutrons. It is thus important to obtain the surface wave characteristic for the safety and the efficiency of system in the IFMIF. In this paper, the free surface oscillation was measured at 175 mm and 15 mm downstream from the nozzle exit with an electro-contact probe apparatus installed on a renewal flow channel of an Osaka University Li loop. At the velocity of more than 9 m/s, maximum wave amplitudes were about 2 mm, which exceed the current design value of IFMIF. However, the number of large-amplitude waves was found to be very low.

Determination and prediction of axial/off-axial mechanical properties of SiC/SiC composites

Nozawa, Takashi; Ozawa, Kazumi; Choi, Y.-B.*; Koyama, Akira*; Tanigawa, Hiroyasu

Fusion Engineering and Design, 87(5-6), p.803 - 807, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.026

A SiC/SiC composite is a candidate material for a demonstration fusion power reactor. Considering the inherent anisotropy of composites with variety of fabric architecture is required to precisely predict axial and off-axial mechanical properties by various failure modes. This study evaluated crack propagation behavior by the various modes to provide a strength anisotropy map and we discussed a methodology to analytically predict this trend. The strength anisotropy maps identified for various fabric orientations clearly indicate that the composites failed by the mixed modes. Specifically, due to the axial anisotropy, five individual modes such as tensile/compressive strengths in the axial/transverse directions, respectively, as well as the in-plane shear strength, are identified to be essential. In this study, with the analytical criterion based on the Tsai-Wu model, the strength anisotropy could satisfactorily be described.

Progress of fusion nuclear technologies in the broader approach framework

Nishitani, Takeo; Garin, P.*; Sugimoto, Masayoshi; Nakajima, Noriyoshi*; Heidinger, R.*; Kimura, Haruyuki; Okano, Kunihiko*; Tobita, Kenji; Yamanishi, Toshihiko; Federici, G.*; et al.

Fusion Engineering and Design, 87(5-6), p.535 - 542, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.01.018

Progress of the fusion nuclear technology in the International Fusion Energy Research Center (IFERC) project and the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project is presented. In the IFERC project, R&D on the blanket materials are progressed. The EU-Japan joint design work on DEMO was initiated in 2011. A high performance computer with 1.3 PFlops is under installation at the Rokkasho BA site, and will be operated from January 2012. In the IFMIF/EVEDA project, the injector of the prototype accelerator was completed and the beam test is on going. The commissioning of the lithium test loop was completed in March 2011, and a lithium flow of 5 m/s was obtained.

Verification test results of a cutting technique for the ITER blanket cooling pipes

Shigematsu, Soichiro; Tanigawa, Hisashi; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Mori, Seiji*; Nakahira, Masataka*; Raffray, R.*; Merola, M.*

Fusion Engineering and Design, 87(7-8), p.1218 - 1223, 2012/08

https://doi.org/10.1016/j.fusengdes.2012.02.108

The current design of the ITER blanket system is a modular configuration and a total of 440 blanket modules are to be installed in the ITER vacuum vessel. Each blanket module consists of the first wall (FW) and the shield block (SB). The FW receives a high heat load from the plasma. The SB shields components from the neutrons generated by the nuclear fusion reaction. The FW will be damaged by the heat load and neutrons, so it requires scheduled replacement. For the FW replacement, cutting/welding tools for the cooling pipes must be able to conduct the following operations: access and cut/weld the pipe from the inside of the cooling pipe. The cutting tool for the pipe end is required to cut flat plate circularly from the surface side of the FW. This paper describes the current status of R&D of the cutting tools for maintenance of the cooling pipe of the FW.