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Miura, Tatsuya*; Nishikata, Atsushi*; Tsuru, Toru*; Yamamoto, Masahiro; Uchiyama, Gunzo; Nojima, Yasuo*; Fujine, Sachio*
Fushoku Boshoku Kyokai Dai-58-Kai Zairyo To Kankyo Toronkai Koenshu, p.15 - 16, 2011/09
There exist some equipments made of Ti or Ti-5Ta alloy in the spent nuclear fuel reprocessing plant. These equipments are investigated for failure by hydrogen embrittlement. To evaluate the stabilities of hydrogen in the solutions using at the reprocessing plant, channel-flow-ellectrode (CFE) method was utilized. The method to determine the dissolved hydrogen concentration was considered from the results of anodic polarization curves.
Tobita, Kenji; Uto, Hiroyasu; Liu, C.; Tanigawa, Hisashi; Tsuru, Daigo; Enoeda, Mikio; Yoshida, Toru; Asakura, Nobuyuki
Fusion Engineering and Design, 85(7-9), p.1342 - 1347, 2010/12
Times Cited Count：35 Percentile：90.77(Nuclear Science & Technology)For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronics and thermal design was carried out to find a blanket concept with reality. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics and Be multipliers. A neutronics-heat coupled analysis determined an optimal arrangement of blanket interior under the constraints of the operating temperature of breeding materials and multipliers. When the cooling water is used under 23 MPa and 290-360 
C, the overall tritium sufficiency is marginally satisfied although blankets with high neutron wall load (
= 5 MW/m
) around the mid-plane do not meet the required local TBR. Based on the results, possible directions for further research are presented.
Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.
JAEA-Research 2010-019, 194 Pages, 2010/08
JAEA-Research-2010-019-01.pdf:48.47MBJAEA-Research-2010-019-02.pdf:19.4MB
This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.
Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.
Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07
Times Cited Count：137 Percentile：97.72(Physics, Fluids & Plasmas)Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).
Tachikawa, Hirokazu*; Kawakubo, Fumie*; Shimizu, Akihiko*; Shibata, Toshio*; Azumi, Kazuhisa*; Inoue, Hiroyuki*; Sugimoto, Katsuhisa*; Tsuru, Toru*; Fujimoto, Shinji*
JAEA-Research 2007-086, 74 Pages, 2008/02
JAEA-Research-2007-086.pdf:5.96MB
The corrosion life time of the overpack has been investigated on the basis of experimental data and past research, assuming the ranging geological environment of Japan. However, some subject for the realization of the overpack design, such as the behavior of the overpack under high pH conditions, the behavior of the overpack with change of near-field environmental condition and the corrosion behavior of the welds have still been left. To take into account these conditions, expert committee composed of metal corrosion science experts were established in the Nuclear Safety Research Association and past research outcomes and the theory of safety assessment for long term corrosion resistance were investigated from the view points of metal corrosion science.
Tachikawa, Hirokazu*; Kawakubo, Fumie*; Shimizu, Akihiko*; Shibata, Toshio*; Sugimoto, Katsuhisa*; Seo, Masahiro*; Tsuru, Toru*; Fujimoto, Shinji*; Inoue, Hiroyuki*
JAEA-Research 2006-058, 80 Pages, 2006/10
JAEA-Research-2006-058.pdf:10.86MB
The Japan Nuclear Cycle Development Institute submitted "Second Progress Report on Research and Development for the Geological Disposal of HLW in Japan" to the Japanese government. This report contains investigations of the corrosion life time of the overpack on the basis of experimental data and past research, assuming the ranging geological environment of Japan. However some subjects, such as the behavior of the overpack under high pH conditions and the behavior of the engineering barrier with change of near-field environmental condition with time for promoting reliability have still been left. To take into account these conditions, expert committee composed of metal corrosion science experts were established in the Nuclear Safety Research Association and past research outcomes and the theory of safety assessment were investigated from the view points of long term stability and corrosion resistance of engineering barrier.
Shibata, Toshio*; *; *; Tsuru, Toru*; Inoue, Hiroyuki*; *
JNC TJ8400 2002-060, 43 Pages, 2003/02
JNC-TJ8400-2002-060.pdf:1.58MB
It is essentially necessary to understand the effect of corrosion products on the corrosion rate of carbon steel in order to evaluate the lifetime of carbon steel overpack, Especially, effect of magnetite on the long term integrity of overpack is one of the important subjects to be solved, because some experimental results showed that the magnetite layer formed on a carbon steel overpack as a corrosion product would accelerates the corrosion rate of the overpack. Various studies have been conducted on the corrosion mechanism of carbon steel in the presence of magnetite, its effect on the overpack lifetime and the countermeasures against the corrosion acceleration. At present, however, the interpretations on the results of these studies are not always consistent each other. In this report, the current status of the studies on corrosion of carbon steel in the presence of magnetite was reviewed, and the unsolved problems and future research subjects were extracted and discussed.
Shibata, Toshio*; *; *; Tsuru, Toru*; Inoue, Hiroyuki*; *
JNC TJ8400 2002-059, 139 Pages, 2003/02
JNC-TJ8400-2002-059.pdf:7.07MB
A technical committee was organized in Japan Society of Corrosion Engineering (JSCE) to review and assess the study of overpack in JNC. The corrosion models for candidate materials for overpacks were developed in terms of corrosion science to contribute the selection of material, establishment of experimental methods and lifetime prediction of overpacks. It is expected that this report is used for the study of overpacks in the process of the research and development of high-level radioactive waste disposal.
Shibata, Toshio*; *; *; Tsuru, Toru*; Inoue, Hiroyuki*
JNC TJ8400 2001-049, 86 Pages, 2002/02
JNC-TJ8400-2001-049.pdf:3.68MB
A technical committee was organized in Japan Society of Corrosion Engineering to review and assess the study of overpacks in JNC. The corrosion models for candidate materials for overpaks were developed in terms of corrosion science to contribute the selection of material, establishment of experimental methods and life prediction of overpacks. It is expected that this report is used for the study of overpacks in the process of the research and development of high-level radioactive waste disposal.
Shibata, Toshio*; *; *; Tsuru, Toru*; Inoue, Hiroyuki*
JNC TJ8400 2001-008, 94 Pages, 2001/02
JNC-TJ8400-2001-008.pdf:2.05MB
A technical committee was organized in Japan Society of Corrosion Engineering to review and assess the study of overpacks in JNC. The corrosion models for candidate materials for overpaks were developed in terms of corrosion science to contribute the selection of material, establishment of experimental methods and life prediction of overpacks. It is expected that this report is used for the study of overpacks in the process of the research and development of high-level radioactive waste disposal.
Shibata, Toshio*; *; *; Tsuru, Toru*; Inoue, Hiroyuki*
JNC TJ8400 2000-013, 38 Pages, 2000/02
JNC-TJ8400-2000-013.pdf:3.25MB
None
Taniguchi, Naoki; ; Kawasaki, Manabu*; Tsuru, Toru*
JNC TN8400 99-003, 88 Pages, 1999/01
Carbon steel is one of the candidate materials for overpacks for high-level radioactive waste disposal in Japan. The estimation of corrosion allowance of carbon steel overpack needs to clarify the type of corrosion and the corrosion rate under repository conditions. The type of the corrosion occuring on overpacks depends on whether carbon steel is passivated or not. If carbon steel is passivated under repository conditions, localized corrosion such as pitting, crevice corrosion and stress corrosion cracking may occur under some conditions. On the other hand, if carbon steel is not passivated under repository conditions, general corrosion will occur. Passivation behavior and corrosion rate of carbon steel were investigated by electrochemical measurements under simulated repository conditions. The results of the measurements showed that carbon steel was hard to passivate in highly compacted bentonite. The immersion tests were carried out in compacted bentonite and average corrosion rates were measured from weight loss and the AC impedance of carbon steel specimens. The database of average corrosion rate were made from the data obtained by the weight loss technique. Based on the database of average corrosion rate in compacted bentonite, the relationship between average corrosion rates and test conditions were investigated. The average corrosion depth for 1000years was also estimated to be less than 5mm. In order to simulate the accumulation of corrosion products after long term, the external current were supplied to carbon steel specimens. After the formation of corrosion products, corrosion rates were measured using AC impedance technique. The results of the measurements showed that the corrosion rate of carbon steel did not increase in the presence of corrosion products formed by external current supply.
Tsujikawa, Shigeo*; *; *; Tsuru, Toru*; Shibata, Toshio*; *
PNC TJ1560 98-001, 164 Pages, 1998/02
None
Tsujikawa, Shigeo*; *; *; Tsuru, Toru*; Shibata, Toshio*; *
PNC TJ1560 97-001, 210 Pages, 1997/03
None
Uchida, Hiroki*; Tada, Eiji*; Tsuru, Toru*; Ishijima, Yasuhiro; Ueno, Fumiyoshi; Yamamoto, Masahiro; Uchiyama, Gunzo; Nojima, Yasuo*; Fujine, Sachio*
no journal, ,
no abstracts in English
Ishijima, Yasuhiro; Motooka, Takafumi; Ueno, Fumiyoshi; Yamamoto, Masahiro; Uchiyama, Gunzo; Sakai, Junichi*; Yokoyama, Kenichi*; Tada, Eiji*; Tsuru, Toru*; Nojima, Yasuo*; et al.
no journal, ,
Titanium and Ti-5mass%Ta alloy has been utilized in nuclear fuel reprocessing plant material because of its superior corrosion resistance in nitric acid solutions. However, Ti alloy have been known to high susceptibility of hydrogen embrittlement. To evaluate properties of hydrogen absorption and hydrogen embrittlement of Ti alloys, cathodic polarization tests and slow strain rate tests (SSRT) under cathodic polarization were carried out. Results show titanium hydrides covered on the surface of metals and hydrides thickness were within 
m. Ti and Ti-5%Ta did not show hydrogen embrittlement by SSRT under cathodic charging. These results suggested that Ti and Ti-5%Ta could absorb hydrogen. But hydrogen did not penetrate inner portion of the metals more than m in depth because titanium hydrides act as barrier of hydrogen diffusion. It is considered that retardation of hydrogen diffusion hindered hydrogen embrittlement of Ti and Ti-5%Ta alloys.
(R=rare earth elements)Yasui, Yukio*; Miyamoto, Takuma*; Kori, Shunsuke*; Terasaki, Ichiro*; Yoshizawa, Daichi*; Mitsuru, Akaki*; Hagiwara, Masayuki*; Matsukawa, Takeshi*; Yoshida, Yukihiko*; Hoshikawa, Akinori*; et al.
no journal, ,
We have measured the specific heat and magnetization up to 50 T of polycrystalline samples of RCrTiO (R=Pr, Nd, Eu, Sm and Gd) and also carried out powder neutron diffraction experiments of NdCrTiO and Nd
PrCrTiO. From the analyses of specific heat data and magnetization curves up to 50 T of the systems, we have evaluated the magnetic interactions between the Cr
- Cr magnetic moments and between the Cr -R ones. The temperature dependence of neutron magnetic reflection intensities of NdCrTiO indicate that both the Cr-moments and the Nd-moments simultaneously order at 
. Basis of the obtained magnetic behavior, the mechanism of magnetoelectric effect of RCrTiO will be discussed.
