Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Loginov, N.*; Mikheyev, A.*; Morozov, V.*; Aksenov, Y.*; Arnoldov, M.*; Berensky, L.*; Fedotovsky, V.*; Chernov, V.*; Nakamura, Hiroo
Journal of Nuclear Materials, 386-388, p.958 - 962, 2009/04
Times Cited Count：6 Percentile：39.83(Materials Science, Multidisciplinary)The IFMIF lithium target mock-ups have been constructed and tested at water and lithium test facilities. Jet velocity in both mock-ups was up to 20 m/s. Calculations and experiments showed instability of lithium flow at conjunction point of straight and concave sections of the mock-up back wall. Profile of water velocity across the mock-up width, jet thickness, and height of waves were measured. A significant increase of thickness of both water and lithium jets near the mock-up side walls was observed. The influence of shape of the nozzle outlet part on jet stability was investigated. Evaporation of lithium from the jet free surface was investigated as well as deposition of lithium on vacuum pipe walls of the target mock-up. It was showed that these phenomena are not so critical for the target efficiency. The possibility of removal of nitrogen in lithium down to 2ppm by means of aluminum getter was showed.
Tsuchiya, Kunihiko; Namekawa, Yoji; Ishida, Takuya
Journal of Nuclear Materials, 386-388, p.1056 - 1059, 2009/04
Times Cited Count：1 Percentile：9.92(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 
Be and Be
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
Fe were observed on the surface of F82H. The thickness of reaction layer between Be-V alloy and F82H was about 10
m. Thus, it has been clarified that compatibility between Be-V alloy and F82H is better than that between Be and F82H.
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：0.00(Materials Science, Multidisciplinary)no abstracts in English
Yamashita, Shinichiro; Yano, Yasuhide; Tachi, Yoshiaki; Akasaka, Naoaki
Journal of Nuclear Materials, 386-388, p.135 - 139, 2009/04
Times Cited Count：14 Percentile：65.35(Materials Science, Multidisciplinary)The heat resistant 11Cr ferritic/martensitic steels were irradiated at 400-670 
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 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 C), the precipitates formation M
C
carbide was primarily dominated. It was a most noticeable microstructural feature that the carbides; MC and MC grew and covered the PABs at this temperature range. (3) Finally, when irradiation temperature was above 650 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.
CTanigawa, Hiroyasu; Klueh, R. L.*; Hashimoto, Naoyuki*; Sokolov, M. A.*
Journal of Nuclear Materials, 386-388, p.231 - 235, 2009/04
Times Cited Count：34 Percentile：88.02(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 MC 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.
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：32 Percentile：87.34(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 
-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.
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：14 Percentile：65.35(Materials Science, Multidisciplinary)no abstracts in English
Miwa, Yukio; Jitsukawa, Shiro; Tsukada, Takashi
Journal of Nuclear Materials, 386-388, p.703 - 707, 2009/04
Times Cited Count：13 Percentile：63.16(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.
Okubo, Nariaki; Miwa, Yukio; Kondo, Keietsu; Kaji, Yoshiyuki
Journal of Nuclear Materials, 386-388, p.290 - 293, 2009/04
Times Cited Count：5 Percentile：29.39(Materials Science, Multidisciplinary)no abstracts in English
10
n/m
Nakano, Junichi; Nemoto, Yoshiyuki; Miwa, Yukio; Usami, Koji; Tsukada, Takashi; Hide, Koichiro*
Journal of Nuclear Materials, 386-388, p.281 - 285, 2009/04
Times Cited Count：5 Percentile：34.78(Materials Science, Multidisciplinary)Crack initiation and crack growth processes of irradiation assisted stress corrosion cracking on stainless steels were studied by slow strain rate testing in oxygenated high temperature water at 561 K. In-situ observation was carried out during SSRT. Specimens of type 304 stainless steel were subjected to a solution annealing (SA), a thermally sensitization (TS), or a cold working (CW) and irradiated to 1.010 n/m (E 
1 MeV) at 323 K in the Japan Material Testing Reactor (JMTR). Crack initiations were observed before the maximum stress would be reached for the CW material in in-situ observation. In fracture surface examination, the TS material exhibited almost intergranular stress corrosion cracking while mixtures of transgranular stress corrosion cracking and ductile dimple fracture were observed for the SA and the CW materials.
Takagi, Ikuji*; Kobayashi, Takashi*; Ueyama, Yutaka*; Moriyama, Hirotake*; Nakamichi, Masaru; Nakamura, Hirofumi; Hayashi, Kimio
Journal of Nuclear Materials, 386-388, p.682 - 684, 2009/04
Times Cited Count：8 Percentile：47.84(Materials Science, Multidisciplinary)no abstracts in English
Kim, H. C.*; Kim, K.*; Lee, Y. S.*; Cho, S. Y.*; Nakajima, Hideo
Journal of Nuclear Materials, 386-388, p.650 - 653, 2009/04
Times Cited Count：6 Percentile：39.83(Materials Science, Multidisciplinary)The characteristics of 316LN welded joints produced by different welding wires were studied through temperature and field dependent magnetization measurements. The magnetic permeability of welded joints showed significant variations depending on the used welding wires. While the self-welded and low-Mn content filler-welded samples showed signs of secondary phase formation, considerable suppression of the secondary phase formation was observed in high-Mn content filler-welded samples. It is concluded that the high-Mn content welding wire is the best one for use in ITER superconducting magnets because its structure is very stable austenite.
Furuya, Kazuyuki*; Ida, Mizuho; Miyashita, Makoto; Nakamura, Hiroo
Journal of Nuclear Materials, 386-388, p.963 - 966, 2009/04
Times Cited Count：19 Percentile：74.83(Materials Science, Multidisciplinary)The current material design of the IFMIF back wall in Japan consists of a stainless steel type-316L and F82H steel. The 316L and F82H are welded each other. The 316L region of the back wall is also welded with the target assembly made of 316L. Since the back-wall is operating under severe neutron irradiation condition (50 dpa/year), it is therefore important to perform metallurgical and mechanical tests for these welds. In result of the tests, significant issues were not found in the F82H/316L TIG-weld. On the other hand, although the 316L/316L YAG-weld offered the weld without any harmful weld defect, the hardness decreased somewhat in the fusion metal. The rupture occurred in the fusion metal, and the strength and elongation decreased somewhat. Furthermore, small dimples include several number of large voids were also seen in the fracture surface.
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
Takaya, Shigeru; Furukawa, Tomohiro; Aoto, Kazumi; M
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：64 Percentile：96.58(Materials Science, Multidisciplinary)The corrosion resistance of ODS steels with 0
3.5 wt% Al and 13.717.3 wt% Cr and of a 12Cr steel were examined. The experiments were conducted at 550 and 650 C up to 3,000 h in stagnant LBE containing 10
and 10
wt% oxygen for the ODS steels and at 550 C up to 5,000 h in stagnant LBE containing 10
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 13.717.3 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 wt% oxygen at 650 C. It is not expected to improve the corrosion resistance by increasing solely Cr content.
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：6 Percentile：39.83(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.
Ando, Masami; Tanigawa, Hiroyasu; Wakai, Eiichi; Stoller, R. E.*
Journal of Nuclear Materials, 386-388, p.315 - 318, 2009/04
Times Cited Count：18 Percentile：73.13(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 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 (
H) obtained by hardness test is almost same level, however radiation hardening (YS) of F82H heat treatment variants is smaller than that of F82H.
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：25 Percentile：81.97(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 LiTiO
and Be (or BeTi). 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 NbAl conductors. The structure and materials of the conducting shell and divertor are also presented.
To, Kentaro; Shikama, Tatsuo*; Katsui, Hirokazu*; Nagata, Shinji*; Tsuchiya, Bun*; Yamauchi, Michinori*; Nishitani, Takeo
Journal of Nuclear Materials, 386-388, p.1027 - 1029, 2009/04
Times Cited Count：5 Percentile：34.78(Materials Science, Multidisciplinary)Fluorescence and long-lasting emission properties of two types of Long-lasting phosphors (LLPs), SrAlO
:Eu
,Dy
and SrAl
O
:Eu,Dy, were examined under 14-MeV neutron irradiation. The fluorescent spectra of the LLPs have characteristic peaks due to their dopants and the fluorescent intensity of SrAlO:Eu,Dy shows good radiation resistance for the neutrons fluence up to 10
n/m. Long-lasting emissions are exhibited after neutron irradiation, and the emission spectra have one peak due to Eu. Further, it is found that there is an optimum neutron fluence at which the longest decay time is obtained.
Hayashi, Takao; Tobita, Kenji; Nakamori, Yuko*; Orimo, Shinichi*
Journal of Nuclear Materials, 386-388, p.119 - 121, 2009/04
Times Cited Count：85 Percentile：98.13(Materials Science, Multidisciplinary)Neutron transport calculations have been carried out to assess the capability of zirconium borohydride (Zr(BH)) and zirconium hydride (ZrH) 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) is lower than ZrH, even though the hydrogen density of Zr(BH) is slightly higher than that of ZrH. High-Z atoms are effective in neutron shielding as well as hydrogen atoms. The combination of steel and Zr(BH) can improve the neutron shielding capability. The combinations of (Zr(BH) + F82H) and (ZrH + F82H) can reduce the thickness of the shield by 6.5% and 19% compared to (water + F82H), respectively. The neutron flux for Zr(BH) 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.
