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Journal Articles

Thermal-stress analysis of IFMIF target back-wall made of reduced-activation ferritic steel and austenitic stainless steel

Ida, Mizuho; Chida, Teruo; Furuya, Kazuyuki*; Wakai, Eiichi; Nakamura, Hiroo; Sugimoto, Masayoshi

Journal of Nuclear Materials, 386-388, p.987 - 990, 2009/04

 Times Cited Count:5 Percentile:37.86(Materials Science, Multidisciplinary)

To clarify IFMIF target back-wall structures and materials with acceptable thermal-stress and deformation due to nuclear heating during the accelerator operation, thermal-stress analysis was done using a code ABAQUS and data of nuclear heating. Two types of back-wall were estimated. One is made of only 316L, and the other is made of 316L at its circumference and F82H a RAF steel at center. Effects of stress-mitigation structure with thickness 2-8 mm, and beam heat loads of 10-100% were estimated. As a result, thermal-stress in the latter back-wall is acceptable level less than 328 MPa for 316L and 455 MPa for F82H even under full heat load, if thickness of the stress-mitigation part is more than 5 mm. On the contrary, thermal-stress in the former is not acceptable. In preliminary tensile tests on dissimilar welding (316L-F82H) specimen, the fracture was occurred in base metal of 316L. Therefore, this welding is expected to be employed as the back-wall.

Journal Articles

Crystal structure of advanced lithium titanate with lithium oxide additives

Hoshino, Tsuyoshi; Sasaki, Kazuya*; Tsuchiya, Kunihiko; Hayashi, Kimio; Suzuki, Akihiro*; Hashimoto, Takuya*; Terai, Takayuki*

Journal of Nuclear Materials, 386-388, p.1098 - 1101, 2009/04

no abstracts in English

Journal Articles

Impact of ceramic coating deposition on the tritium permeation in the Japanese ITER-TBM

Nakamichi, Masaru; Nakamura, Hirofumi; Hayashi, Kimio; Takagi, Ikuji*

Journal of Nuclear Materials, 386-388, p.692 - 695, 2009/04

 Times Cited Count:10 Percentile:59.24(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Recent findings on blistering and deuterium retention in tungsten exposed to high-fluence deuterium plasma

Shu, Wataru; Kawasuso, Atsuo; Yamanishi, Toshihiko

Journal of Nuclear Materials, 386-388, p.356 - 359, 2009/04

 Times Cited Count:34 Percentile:91.16(Materials Science, Multidisciplinary)

Tungsten is a most promising plasma facing material because of its high melting point. The blistering and deuterium retention in recrystallized tungsten samples were investigated by using simulated edge-plasma of fusion reactors. High-dome blisters appeared at the tungsten surface after deuterium plasma exposure, and their ratios of height against chord of the blisters were even one-order greater that reported before. In addition, there was a cavity in the inside of small blisters, whereas there was a void/crack along the grain boundary beneath the big blister and there is no lid for big blisters. Besides the strong dependence upon the exposure temperature, blistering and deuterium retention also showed significant dependence upon the features of microstructure.

Journal Articles

Effect of two-steps heat treatments on irradiation hardening in F82H irradiated at 573 K

Ando, Masami; Tanigawa, Hiroyasu; Wakai, Eiichi; Stoller, R. E.*

Journal of Nuclear Materials, 386-388, p.315 - 318, 2009/04

 Times Cited Count:16 Percentile:73.65(Materials Science, Multidisciplinary)

Radiation hardening and embrittlement due to neutron irradiation around 573 K are the important issues on RAF/M steels. It is expected that the improvement of radiation hardening might be one of effective ways to control the mechanical properties after irradiation. The purposes of this study are to find the condition of heat treatment for minimum of radiation hardening in F82H steel using Neutron/Ion-irradiation and to examine a correlation between tensile property and micro-hardness before/after irradiation. Neutron irradiation was performed in HFIR up to 9 dpa. Ion-irradiation at $$sim$$573 K was carried out at the TIARA facility of JAEA. For the results of tensile test and hardness test of F82H and F82H heat treatment variants neutron-irradiated at 573 K, all specimens caused radiation hardening. The radiation hardening ($$Delta$$H) obtained by hardness test is almost same level, however radiation hardening ($$Delta$$YS) of F82H heat treatment variants is smaller than that of F82H.

Journal Articles

Compatibility between Be-V alloy and F82H steel

Tsuchiya, Kunihiko; Namekawa, Yoji; Ishida, Takuya

Journal of Nuclear Materials, 386-388, p.1056 - 1059, 2009/04

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

Beryllium alloys such as Be-Ti and Be-V have been expected as promising candidates for advanced neutron multipliers of DEMO blankets from viewpoints of high melting point, high beryllium content, low radio activation, good chemical stability. In this study, the compatibility between Be-V alloy and F82H was investigated. The Be-7at%V specimens that includes $$alpha$$Be and Be$$_{12}$$V phases were tested. From the XRD analysis of the specimens after annealing of the Be-V alloy in contact with F82H, reaction products such as BeNi and Be$$_{2}$$Fe were observed on the surface of F82H. The thickness of reaction layer between Be-V alloy and F82H was about 10$$mu$$m. Thus, it has been clarified that compatibility between Be-V alloy and F82H is better than that between Be and F82H.

Journal Articles

Preliminary test for reprocessing technology development of tritium breeders

Hoshino, Tsuyoshi; Tsuchiya, Kunihiko; Hayashi, Kimio; Nakamura, Mutsumi*; Terunuma, Hitoshi*; Tatenuma, Katsuyoshi*

Journal of Nuclear Materials, 386-388, p.1107 - 1110, 2009/04

 Times Cited Count:9 Percentile:55.98(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Design and trial fabrication of a dismantling apparatus for irradiation capsules of solid tritium breeder materials

Hayashi, Kimio; Nakagawa, Tetsuya; Onose, Shoji; Ishida, Takuya; Nakamichi, Masaru; Takatsu, Hideyuki; Nakamura, Mutsumi*; Noguchi, Tsuneyuki*

Journal of Nuclear Materials, 386-388, p.1083 - 1086, 2009/04

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Torus configuration and materials selection on a fusion DEMO reactor, SlimCS

Tobita, Kenji; Nishio, Satoshi; Tanigawa, Hiroyasu; Enoeda, Mikio; Isono, Takaaki; Nakamura, Hirofumi; Tsuru, Daigo; Suzuki, Satoshi; Hayashi, Takao; Tsuchiya, Kunihiko; et al.

Journal of Nuclear Materials, 386-388, p.888 - 892, 2009/04

 Times Cited Count:24 Percentile:84.45(Materials Science, Multidisciplinary)

SlimCS is the conceptual design of a compact fusion DEMO plant assuming technologies foreseeable in 2020s-2030s. Considering continuity of blanket technology from the Japanese proposal on ITER-TBM, the prime option of blanket is water-cooled solid breeder with Li$$_{2}$$TiO$$_{3}$$ and Be (or Be$$_{12}$$Ti). A reduced-activation ferritic-martensitic steel and pressurized water are chosen as the structural material and coolant, respectively. Toroidal coils produce the peak magnetic field above 16 T using the RHQT processed Nb$$_{3}$$Al conductors. The structure and materials of the conducting shell and divertor are also presented.

Journal Articles

Recent advances and issues in development of silicon carbide composites for fusion applications

Nozawa, Takashi; Hinoki, Tatsuya*; Hasegawa, Akira*; Koyama, Akira*; Kato, Yutai*; Snead, L. L.*; Henager, C. H. Jr.*; Hegeman, J. B. J.*

Journal of Nuclear Materials, 386-388, p.622 - 627, 2009/04

 Times Cited Count:110 Percentile:99.31(Materials Science, Multidisciplinary)

A new class SiC/SiC composite have been developed for fusion. While the development efforts has resulted in a radiation-resistant SiC/SiC, development continues to improve on engineering properties of this composite. With the completion of the "proof-of-principal" phase, the R&D on SiC/SiC is now shifting to the more pragmatic phase of material data-basing. Critical application issues still remain, including (1) heavy irradiation effect with considerations of He/H synergistic effects and irradiation creep, (2) erosion-corrosion behavior, (3) nuclear transmutation other than He/H, and (4) joining. Understanding the irradiation effect on electrical conductivity is another important issue for the FCI application. Along with the review in material development, characterization, and irradiation effect studies, this paper provides an introductive work toward standardization of the test methodology.

Journal Articles

Fusion materials development program in the Broader Approach activities

Nishitani, Takeo; Tanigawa, Hiroyasu; Jitsukawa, Shiro; Nozawa, Takashi; Hayashi, Kimio; Yamanishi, Toshihiko; Tsuchiya, Kunihiko; M$"o$slang, A.*; Baluc, N.*; Pizzuto, A.*; et al.

Journal of Nuclear Materials, 386-388, p.405 - 410, 2009/04

 Times Cited Count:31 Percentile:89.67(Materials Science, Multidisciplinary)

The establishment of the breeding blanket technology is one of the most important engineering issues on the DEMO development. For the DEMO blanket, developments of the structural materials and functional materials such as tritium breeder and neutron multiplier. Which should be used under the savior circumstance such as high neutron fluence, high temperature and strong magnetic field, are urgent issues. In the Broader Approach activities initiated by EU and Japan, developments of reduced activation ferritic martensitic steels as a DEMO blanket structural material, SiC/SiC composites, advanced tritium breeders and neutron multiplier for DEMO blankets, are planed as common interest issues of EU and Japan. This paper describes the overview of the development program.

Journal Articles

Effect of high dose/high temperature irradiation on the microstructure of heat resistant 11Cr ferritic/martensitic steels

Yamashita, Shinichiro; Yano, Yasuhide; Tachi, Yoshiaki; Akasaka, Naoaki

Journal of Nuclear Materials, 386-388, p.135 - 139, 2009/04

 Times Cited Count:12 Percentile:62.13(Materials Science, Multidisciplinary)

The heat resistant 11Cr ferritic/martensitic steels were irradiated at 400-670 $$^{circ}$$C up to 100 dpa in FFTF and JOYO. The microstructures of unirradiated 11Cr ferritic/martensitic steels consist of laths, dislocation, and carbide. Almost of the prior austenitic boundaries (PABs) were partially decorated with carbides. It was observed from the results of post irradiation microstructural examinations that the irradiation-induced microstructures were classified into the following three types depending on irradiation temperature; (1) When irradiated at 400-450 $$^{circ}$$C, both dislocation loops and cavities with less than 30 nm in diameter were formed in the ferrite phase. On the other hand, the void swelling was about 0.05%. (2) In the case of irradiation at moderate temperature (500-600 $$^{circ}$$C), the precipitates formation M$$_{23}$$C$$_{6}$$ carbide was primarily dominated. It was a most noticeable microstructural feature that the carbides; M$$_{23}$$C$$_{6}$$ and M$$_{6}$$C grew and covered the PABs at this temperature range. (3) Finally, when irradiation temperature was above 650 $$^{circ}$$C microstructures were drastically-changed. Microstructural observations revealed that formation and growth of equi-axial grain occurred in addition to recovery of laths, growth of carbides simultaneously at high temperature. This remarkable microstructural change might be closely related to a severe degradation in the mechanical properties.

Journal Articles

Advanced neutron shielding material using zirconium borohydride and zirconium hydride

Hayashi, Takao; Tobita, Kenji; Nakamori, Yuko*; Orimo, Shinichi*

Journal of Nuclear Materials, 386-388, p.119 - 121, 2009/04

 Times Cited Count:70 Percentile:97.93(Materials Science, Multidisciplinary)

Neutron transport calculations have been carried out to assess the capability of zirconium borohydride (Zr(BH$$_{4}$$)$$_{4}$$) and zirconium hydride (ZrH$$_{2}$$) as advanced shield materials, because excellent shields can be used to protect outer structural materials from serious activation. The neutron shielding capability of Zr(BH$$_{4}$$)$$_{4}$$ is lower than ZrH$$_{2}$$, even though the hydrogen density of Zr(BH$$_{4}$$)$$_{4}$$ is slightly higher than that of ZrH$$_{2}$$. High-Z atoms are effective in neutron shielding as well as hydrogen atoms. The combination of steel and Zr(BH$$_{4}$$)$$_{4}$$ can improve the neutron shielding capability. The combinations of (Zr(BH$$_{4}$$)$$_{4}$$ + F82H) and (ZrH$$_{2}$$ + F82H) can reduce the thickness of the shield by 6.5% and 19% compared to (water + F82H), respectively. The neutron flux for Zr(BH$$_{4}$$)$$_{4}$$ is drastically reduced in the range of neutron energy below 100 eV compared to other materials, due to the effect of boron, which can lead to a reduction of radwaste from fusion reactors.

Journal Articles

Effects of aluminum on high-temperature strength of 9Cr-ODS steel

Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki; Asayama, Tai; Kim, S.-W.; Ukai, Shigeharu*; Narita, Takeshi*; Sakasegawa, Hideo*

Journal of Nuclear Materials, 386-388, p.479 - 482, 2009/04

 Times Cited Count:18 Percentile:77.39(Materials Science, Multidisciplinary)

This paper discusses the effects of small portion of Al contamination ($$<$$0.1wt%) on the high-temperature strength and microstructure of 9Cr-ODS steel. Increasing Al concentration is shown to provide small reduction of ultimate tensile strength as well as 0.2% proof stress at 973 K and 1073 K accompanied by reduction of elongated grains i.e. residual-$$alpha$$ ferrite acting as reinforcement phase. Addition of Al appears to increase the proportion of ferrite phase, which is contrary to general behavior in conventional steels. This unique behavior could be peculiar to the non-equilibrium materials such as mechanically-alloyed alloy. Computer simulation on phase transformation suggests that the fine oxide dispersion in the elongated ferrite could be attributable to the preferential partitioning of Ti and W in ferrite than in austenite at hot-extrusion process at 1423 K.

Journal Articles

Cyclically induced softening in reduced activation ferritic/martensitic steel before and after neutron irradiation

Kim, S.-W.; Tanigawa, Hiroyasu; Hirose, Takanori; Koyama, Akira*

Journal of Nuclear Materials, 386-388, p.529 - 532, 2009/04

 Times Cited Count:12 Percentile:64.63(Materials Science, Multidisciplinary)

Low cycle fatigue (LCF) results at ambient temperature under diametral strain controlled conditions of the reduced activation ferritic/martensitic steel, F82H IEA heat before and after neutron irradiated samples are reported. The results show that cyclic softening behavior is the main mechanical feature observed in this material. A detailed analysis for the hysteresis loops was carried out to determine the friction and back stresses. The friction stress is equivalent to the resistance which the dislocations have to overcome to keep moving in the lattice. The back stress depends on the density of long-range impenetrable obstacles that are created by the dislocations movement such as pile-ups. The cyclic softening of F82H IEA heat is related with the decrease of the friction stress. Moreover, the friction and back stress behavior of neutron irradiated samples show significantly different behavior than those of unirradiated samples.

Journal Articles

Corrosion behavior of Al-alloying high Cr-ODS steels in lead-bismuth eutectic

Takaya, Shigeru; Furukawa, Tomohiro; Aoto, Kazumi; M$"u$ller, G.*; Weisenburger, A.*; Heinzel, A.*; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; et al.

Journal of Nuclear Materials, 386-388, p.507 - 510, 2009/04

 Times Cited Count:46 Percentile:94.98(Materials Science, Multidisciplinary)

The corrosion resistance of ODS steels with 0$$sim$$3.5 wt% Al and 13.7$$sim$$17.3 wt% Cr and of a 12Cr steel were examined. The experiments were conducted at 550 and 650 $$^{circ}$$C up to 3,000 h in stagnant LBE containing 10$$^{-6}$$ and 10$$^{-8} $$wt% oxygen for the ODS steels and at 550 $$^{circ}$$C up to 5,000 h in stagnant LBE containing 10$$^{-8}$$ wt% oxygen for the 12Cr steel, respectively. Protective Al oxide scales were formed on the surfaces of ODS steels with about 3.5 wt% Al and 13.7$$sim$$17.3 wt% Cr. The addition of Al is very effective to improve the corrosion resistance of ODS steels. The ODS steel with 16 wt% Cr and no Al does not show any corrosion resistance except for the specimen exposed to LBE with 10$$^{-6}$$ wt% oxygen at 650 $$^{circ}$$C. It is not expected to improve the corrosion resistance by increasing solely Cr content.

Journal Articles

Hardening mechanisms of reduced activation ferritic/martensitic steels irradiated at 300 $$^{circ}$$C

Tanigawa, Hiroyasu; Klueh, R. L.*; Hashimoto, Naoyuki*; Sokolov, M. A.*

Journal of Nuclear Materials, 386-388, p.231 - 235, 2009/04

 Times Cited Count:26 Percentile:86.28(Materials Science, Multidisciplinary)

It has been reported that reduced-activation ferritic/martensitic steels (RAFMs) showed a variety of changes in ductile-brittle transition temperature and yield stress after irradiation at 573 K up to 5 dpa, and those differences could not be interpreted solely by the difference of dislocation microstructure induced by irradiation. To investigate the impact of other microstructural feature, i.e. precipitates, the precipitation behavior of F82H, ORNL 9Cr-2WVTa, and JLF-1 was examined. It was revealed that irradiation-induced precipitation and amorphization of precipitates partly occurred and caused the different precipitation on block, packet and prior austenitic grain boundaries. In addition to these phenomena, irradiation-induced nano-size precipitates were also observed in the matrix. It was also revealed that the chemical compositions of precipitates approached the calculated thermal equilibrium state of M$$_{23}$$C$$_{6}$$ at an irradiation temperature of 573 K. Over all, these observations suggests that the variety of embrittlement and hardening of RAFMs observed at 573 K irradiation up to 5 dpa might be the consequence of the transition phenomena that occur as the microstructure approaches thermal equilibrium during irradiation at 573 K.

Journal Articles

In-pile creep rupture properties of ODS ferritic steel claddings

Kaito, Takeji; Otsuka, Satoshi; Inoue, Masaki; Asayama, Tai; Uwaba, Tomoyuki; Mizuta, Shunji; Ukai, Shigeharu*; Furukawa, Tomohiro; Ito, Chikara; Kagota, Eiichi; et al.

Journal of Nuclear Materials, 386-388, p.294 - 298, 2009/04

 Times Cited Count:29 Percentile:88.31(Materials Science, Multidisciplinary)

In order to examine irradiation effect on creep rupture strength of Oxide Dispersion Strengthened (ODS) steel claddings, the in-pile creep rupture test was conducted using Material Testing Rig with Temperature Control (MARICO)-2 in the experimental fast reactor JOYO. Fourteen creep rupture events were successfully detected by the temperature change in each capsule and the $$gamma$$-ray spectrometry of the cover gas. Time to creep ruptures of six ODS steel specimens were identified by means of Laser Resonance Ionization Mass Spectrometry (RIMS), and no irradiation effect on creep rupture strength was confirmed within the irradiation condition in the MARICO-2 test.

Journal Articles

Stress corrosion cracking susceptibility of a reduced-activation martensitic steel F82H

Miwa, Yukio; Jitsukawa, Shiro; Tsukada, Takashi

Journal of Nuclear Materials, 386-388, p.703 - 707, 2009/04

 Times Cited Count:9 Percentile:55.98(Materials Science, Multidisciplinary)

In order to examine the stress corrosion cracking (SCC) susceptibility of reduced activation ferritic/martensitic steel, F82H, slow strain rate test (SSRT) was performed at various temperature in oxygenated or hydrogenated water. Test specimens of F82H were heat-treated at various temperature conditions, or were cold-worked to imitate radiation hardening and machined to make single edge notch, or were neutron-irradiated at 493 K to 3.4 dpa. It was found that in unirradiated specimen, IGSCC occurred when specimen was normalized only, and TGSCC occurred when cold-worked (over 23%) and notched specimen was tested by SSRT at 573 K in oxygenated water. In irradiated specimen, TGSCC occurred, when SSRT was conducted at 573 K in hydrogenated (DH = 1 ppm) water or when the notched specimen was tested by SSRT at 573 K in oxygenated (DO = 10 ppm) water.

Journal Articles

Damage process and luminescent characteristics in silica glasses under ion irradiation

Nagata, Shinji*; Katsui, Hirokazu*; Tsuchiya, Bun*; Inoue, Aichi; Yamamoto, Shunya; To, Kentaro; Shikama, Tatsuo*

Journal of Nuclear Materials, 386-388, p.1045 - 1048, 2009/04

 Times Cited Count:13 Percentile:67.22(Materials Science, Multidisciplinary)

no abstracts in English

36 (Records 1-20 displayed on this page)