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Journal Articles

Atomistic weak interaction criterion for the specificity of liquid metal embrittlement

Yamaguchi, Masatake; Tsuru, Tomohito; Itakura, Mitsuhiro; Abe, Eiji*

Scientific Reports (Internet), 12(1), p.10886_1 - 10886_7, 2022/07

 Times Cited Count:0 Percentile:0(Multidisciplinary Sciences)

no abstracts in English

Journal Articles

Effect of crystal orientation on incipient plasticity during nanoindentation of magnesium

Somekawa, Hidetoshi*; Tsuru, Tomohito; Singh, A.*; Miura, Seiji*; Schuh, C. A.*

Acta Materialia, 139, p.21 - 29, 2017/10

 Times Cited Count:26 Percentile:80.35(Materials Science, Multidisciplinary)

The effect of crystal orientation on incipient plasticity during nanoindentation was investigated by experiments and molecular statics simulation. Pop-in behavior is a result of dislocation activity, and is therefore influenced by crystal orientation. Experimental results using single crystals indicated that indentations on the basal plane had higher pop-in loads and larger pop-in displacements than those on the prismatic plane, an effect also captured by molecular statics simulation. The difference can be traced to the types of activated dislocations, with not only basal but also pyramidal dislocations active for indentations on the basal plane, but only basal dislocations triggered at the first pop-in on the prismatic plane.

Journal Articles

Neutron diffraction study on anisotropy of strain age hardening in ferritic steel

Suzuki, Tetsuya*; Yamanaka, Keisuke*; Ishino, Mayuko*; Shinohara, Yasuhiro*; Nagai, Kensuke*; Tsuru, Eiji*; Xu, P. G.

Tetsu To Hagane, 98(6), p.262 - 266, 2012/06

Journal Articles

Compact DEMO, SlimCS; Design progress and issues

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:135 Percentile:97.7(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$$^{2}$$ 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).

Journal Articles

Development of water-cooled solid breeder test blanket module in JAEA

Akiba, Masato; Enoeda, Mikio; Tsuru, Daigo; Tanigawa, Hisashi; Hirose, Takanori; Mori, Kensuke*; Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Nishi, Hiroshi; et al.

Fusion Engineering and Design, 84(2-6), p.329 - 332, 2009/02

 Times Cited Count:16 Percentile:71.26(Nuclear Science & Technology)

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as Test Blanket Module (TBM). JAEA has been extensively developing a Water-Cooled Solid Breeder Test Blanket Module (WCSB TBM) for ITER. This paper describes results of recent R&D activities on WCSB TBM in JAEA. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by Hot Isostatic Press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 $$^{circ}$$C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m$$^{2}$$, 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.

Oral presentation

Numerical simulation of helium purge gas with tritium transfer in pebble bed for WCSB test blanket module

Seki, Yohji; Akiba, Masato; Enoeda, Mikio; Suzuki, Satoshi; Nishi, Hiroshi; Ezato, Koichiro; Yokoyama, Kenji; Tanigawa, Hisashi; Mori, Seiji; Tanzawa, Sadamitsu; et al.

no journal, , 

The one-dimensional nuclear and thermal analyses on Test Blanket Module (TBM) in ITER have been performed for emphasizing on optimized layer structures of a ceramic tritium breeder ($$Li_{2}TiO_{3}$$) and a beryllium neutron multiplier $$Be$$. The numerical simulation of the helium purge gas in the breeder pebble bed also has been performed. The main results of our study are as follows: (1) In the case of the single packing for multiplier pebble bed, The tritium product ratio of single packing is comparable in magnitude to that of binary packing by setting the two layers of Be behind a layer of $$Li_{2}TiO_{3}$$. (2) The high concentration of the tritium stays near the first wall and membrane panel because of the effect of insufficient convective diffusion in low Reynolds number. This work contributes to the designs of the TBM and demonstration blanket.

Oral presentation

Estimation of strain aging for ferritic steel by neutron diffraction

Suzuki, Tetsuya*; Yamanaka, Keisuke*; Nagai, Kensuke*; Tsuru, Eiji*; Xu, P. G.; Suzuki, Hiroshi

no journal, , 

no abstracts in English

Oral presentation

Evaluation of the hydrogen embrittlement susceptibility for pure titanium under cathodic charging and observation of the crack initiation and propagation

Uchida, Hiroki*; Tada, Eiji*; Tsuru, Toru*; Ishijima, Yasuhiro; Ueno, Fumiyoshi; Yamamoto, Masahiro; Uchiyama, Gunzo; Nojima, Yasuo*; Fujine, Sachio*

no journal, , 

no abstracts in English

Oral presentation

Hydrogen absorption behavior of titanium alloys by cathodic polarization

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 $$mu$$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 $$mu$$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.

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