Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Kogawara, Takafumi; Wakai, Eiichi; Kikuchi, Takayuki; Yamamoto, Michiyoshi; Molla, J.*
Fusion Engineering and Design, 86(12), p.2904 - 2907, 2011/12
Times Cited Count：3 Percentile：24.29(Nuclear Science & Technology)no abstracts in English
Shiba, Kiyoyuki; Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro
Fusion Engineering and Design, 86(12), p.2895 - 2899, 2011/12
Times Cited Count：45 Percentile：93.86(Nuclear Science & Technology)Aging properties of reduced activation ferritic/martensitic steel F82H was researched at temperature ranging from 400
C to 650C up to 100,000 hr. Microstructure, tensile, and Charpy properties were carried out. Laves was found at temperatures between 550 and 650C and M
C carbides were found at the temperatures between 500 and 600C over 10,000 hr. These precipitates caused degradation in toughness, especially at temperatures ranging from 550C to 650C. Tensile properties do not have serious aging effect, except for 650C, which caused large softening even after 10.000 hr. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in DBTT caused by the large Laves phase precipitation at grain boundary was observed in the 650C aging. Laves precipitates at grain boundary also degrades the USE of the aged materials. These aging test results provide F82H can be used up to 30,000 hr at 550C.
Nishitani, Takeo; Yamanishi, Toshihiko; Tanigawa, Hiroyasu; Nozawa, Takashi; Nakamichi, Masaru; Hoshino, Tsuyoshi; Koyama, Akira*; Kimura, Akihiko*; Hinoki, Tatsuya*; Shikama, Tatsuo*
Fusion Engineering and Design, 86(12), p.2924 - 2927, 2011/12
Times Cited Count：7 Percentile：46.23(Nuclear Science & Technology)Several technical R&D activities related to the blanket materials are newly launched as a part of the Broader Approach (BA) activities, which was initiated by the EU and Japan. According to the common interests of these parties for DEMO, R&Ds on reduced activation ferritic/martensitic (RAFM) steels as structural material, SiCf/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out through the BA DEMO R&D program, in order to establish the technical bases on the blanket materials and the tritium technology required for DEMO design. This paper describes overall schedule of those R&D activities and recent progress in Japan carried out by JAEA as the domestic implementing agency on BA, collaborating with Japanese universities and other research institutes.
Chida, Yutaka; Iguchi, Masahide; Takano, Katsutoshi; Nakajima, Hideo; Osemochi, Koichi*; Niimi, Kenichiro*; Tokai, Daisuke*; Gallix, R.*
Fusion Engineering and Design, 86(12), p.2900 - 2903, 2011/12
Times Cited Count：10 Percentile：57.82(Nuclear Science & Technology)TF coil structures, which support large electromagnetic force generated in TF coils under the cryogenic temperature (about 4K), are the mega welding structures composed of coil case and support structures made of high strength and high toughness stainless steel. JAEA started the study on welding trials for heavy thickness materials since 2008 and is planning of full scale mock-up model fabrication for main sub-components (1 set of inboard side and 1set of outboard side) in 2010 in order to investigate the technical issues for manufacturing of TF coil structures. This paper introduces the results on welding trials and status of full scale mock-up model fabrication to confirm the validity of welding technology and manufacturing design before fabricating actual products.
Someya, Yoji; Takase, Haruhiko; Uto, Hiroyasu; Tobita, Kenji; Liu, C.; Asakura, Nobuyuki
Fusion Engineering and Design, 86(9-11), p.2269 - 2272, 2011/11
Times Cited Count：37 Percentile：91.56(Nuclear Science & Technology)Conceptual design of a tritium-breeding blanket for SlimCS has been studied. The blanket structure with neutron multiplier Be-plate was designed to be as thin as possible with keeping high Tritium Breeding Ratio (TBR). However, a structure of the blanket is complexity and the manufacture of the blanket is difficult from the viewpoint of engineering. Therefore, simplification of blanket structure is necessary for SlimCS. In this paper, we propose a simple blanket structure without decreasing the net TBR below 1.05. The proposed blanket structure is mixed Li
SiO pebbles or Li
O pebbles for the tritium breeding and Be
Ti pebbles for the neutron multiplication and these pebbles are filled in the blanket. ANIHEAT code with the nuclear data library FENDL-2.0 was used for the calculations of the neutronics and thermal analyses. As a result, it is shown that LiO pebbles blanket mixed with BeTi pebbles is the most effective and the TBR is greater than 1.05.
Iwai, Yasunori; Sato, Katsumi; Yamanishi, Toshihiko
Fusion Engineering and Design, 86(9-11), p.2164 - 2167, 2011/10
Times Cited Count：19 Percentile：77.89(Nuclear Science & Technology)We have developed the hydrophobic Pt catalysts applicable for tritium oxidation in the presence of saturated water vapor at room temperature. A new type of hydrophobic catalyst, Pt/ASDBC, has been prepared by dipositing platinum on alkyl-styrene diviyl-benzene copolymer (ASDBC). Pt/ASDBC is more hydrophobic than Pt/SDBC that is a promising catalyst for the water detritiation system. The deposited platinum used to prepare Pt/ASDBC catalyst was 1.0 g/L. The value was approximately half of a commercial Japanese Pt/SDBC catalyst. Tritium oxidation tests of the catalysts using 3 GBq/m
of tritium were performed in the absence/presence of saturated water vapor at room temperature. Tritium oxidation sufficient for room temperature recombiner was demonstrated using Pt/ASDBC catalyst.
Yamanishi, Toshihiko; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Hara, Masanori*; Sugiyama, Takahiko*; Okuno, Kenji*
Fusion Engineering and Design, 86(9-11), p.2152 - 2155, 2011/10
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)It is quite significant subject how to confine the tritium in a fusion reactor. Especially, it is strongly desired to get the data for tritiated water. This is because tritiated water is much hazardous than the hydrogen form of tritium. As for the behavior of high concentration tritium water, we could get a series of valuable data for the corrosion of the tritiated water against metal materials. In the case where a metal material is in water, an oxidized layer is formed at the surface of the metal. The oxidized layer functions as a passive layer for the corrosion. However, it has been observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). The chemical exchange column has been applied in ITER as the tritium recovery system from tritiated water. A set of data for an advanced chemical exchange column has been obtained.
Hoshino, Tsuyoshi; Terai, Takayuki*
Fusion Engineering and Design, 86(9-11), p.2168 - 2171, 2011/10
Times Cited Count：62 Percentile：96.72(Nuclear Science & Technology)The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 (
Li) in tritium breeding materials. However, natural Li contains only about 7.6 at.% Li. In Japan, new lithium isotope separation technique using ionic-liquid impregnated organic membranes have been developed. The improvement in the durability of the ionic-liquid impregnated organic membrane is one of the main issues for stable, long-term operation of electrodialysis cells while maintaining good performance. Therefore, we developed highly-durable ionic-liquid impregnated organic membrane. Both ends of the ionic-liquid impregnated organic membrane were covered by a nafion 324 overcoat to prevent the outflow of the ionic liquid. The transmission of Lithium aqueous solution after 10 hours under the highly-durable ionic-liquid impregnated organic membrane is almost 13%. So this highly-durable ionic-liquid impregnated organic membrane for long operating of electrodialysis cells has been developed through successful prevention of ion liquid dissolution.
Ishikawa, Masao; Kawano, Yasunori; Imazawa, Ryota; Sato, Satoshi; Vayakis, G.*; Bertalot, L.*; Yatsuka, Eiichi; Hatae, Takaki; Kondoh, Takashi; Kusama, Yoshinori
Fusion Engineering and Design, 86(6-8), p.1286 - 1289, 2011/10
Times Cited Count：1 Percentile：10.08(Nuclear Science & Technology)The nuclear heating rates of the optical mirrors of the poloidal polarimeter installed in the equatorial port plug of ITER are calculated. Since the system cannot have a sufficiently labyrinthine structure and the second mirrors are located nearly as close to the plasma as the first mirrors due to limited space, the nuclear heating rate of the second mirrors is as high as that of the first mirrors. However, it is possible to reduce the nuclear heating rates of the mirrors if the blanket shield module provides a sufficient degree of neutron shielding.
Tobita, Kenji; Uto, Hiroyasu; Kakudate, Satoshi; Takase, Haruhiko; Asakura, Nobuyuki; Someya, Yoji; Liu, C.
Fusion Engineering and Design, 86(9-11), p.2730 - 2734, 2011/10
Times Cited Count：13 Percentile：66.72(Nuclear Science & Technology)For high availability of DEMO operation, sector horizontal transport hot cell maintenance scheme was studied. Transport of sector with 730 tons is carried out using a wheeled platform. The driving force of pulling the sector into a cask is ball screws. The fulcrum of the ball screws is the cryostat wall so that a large pulling force is expected with no-counter balance. The cask containing the sector is delivered by air casters from the cryostat to the hot cell. For the maintenance scheme, new concepts such as transfer of the tilting forces of toroidal coils using ropes and shafts and supports for the tilting force using reinforced concrete floor or cryostat wall were proposed. Based on the maintenance concept, the period required for replacement of all sectors is estimated to be 35.5-67.5 days, satisfying the design target (shorter than 3 months).
Kobayashi, Takayuki; Isayama, Akihiko; Hasegawa, Koichi; Suzuki, Sadaaki; Hiranai, Shinichi; Sato, Fumiaki; Wada, Kenji; Yokokura, Kenji; Shimono, Mitsugu; Sawahata, Masayuki; et al.
Fusion Engineering and Design, 86(6-8), p.763 - 767, 2011/10
Times Cited Count：6 Percentile：41.54(Nuclear Science & Technology)Progress of antenna development of the Electron Cyclotron Range of Frequency system for JT-60 SA is presented. Capability of pulse length of 100 s, which requires active cooling for mirrors, and flexibility of beam injection angles in both poloidal and toroidal directions are required for the antenna with high reliability. Mechanical and structural design works of the launcher (antenna and its support with steering structure) based on a linear motion antenna concept are in progress. The key component is a long-stroke bellows which enables to alter poloidal injection angle and a bellows which enables to alter toroidal injection angle. Using a newly fabricated mock-up of the steering structure, it was confirmed that the antenna was mechanically realized for poloidal and toroidal injection angle ranges of -10 to +45 and -15 to +15, respectively. Those angles are consistent with angles required in JT-60SA. The results of thermal and structural analyses are also presented.
Kwak, J. G.*; Wang, S. J.*; Bae, Y. D.*; Kim, S. H.*; Hwang, C. K.*; Moriyama, Shinichi
Fusion Engineering and Design, 86(6-8), p.938 - 941, 2011/10
Times Cited Count：1 Percentile：10.08(Nuclear Science & Technology)KAERI have been developing the transmitters for ICRF heating for KSTAR and the cyclotron accelerator since 1996. The toroidal magnetic field of KSTAR is nominally 3 T so that 25-60 MHz transmitter is required to cover ICRF heating scenarios of the KSTAR. The first transmitter is operating up to 60 MHz and it succeeded in achieving 2 MW for 300 s in 2008. Up to 300 kW RF power was successfully injected to KSTAR plasmas. The second one is 70 kW/CW transmitter used for the cyclotron accelerator and their frequency range is from 25 to 50 MHz. Its engineering design was finished. The third one is 1 MW/VHF transmitter which was loaned from JAEA. As the operating ICRF frequency of KSTAR is lower, its cavity structure will be modified from 110 MHz to 60 MHz. The test results of 60 MHz and lessons from the high power test of 2 MW transmitter will be introduced and the circuit analysis and engineering design work for the second and third amplifiers will be shown.
Sakasegawa, Hideo; Tanigawa, Hiroyasu; Kano, Sho; Enomoto, Masato*
Fusion Engineering and Design, 86(9-11), p.2541 - 2544, 2011/10
Times Cited Count：17 Percentile：75.07(Nuclear Science & Technology)Reduced Activation Ferritic/Martensitic steels (RAFMs) are leading candidates for the structural material of DEMO blanket module. Through the Broader Approach (BA) activity in Japan, the fabrication techniques for the DEMO blanket module has been studied and developed. In the techniques, the development of joining technique is especially important for fabricating the complicated structure of blanket module. In particular, Hot Isostatic Pressing (HIP) has been applied to joining cooling channels with a rectangular cross section. During and after HIP, the structural material are exposed to various heat treatments such as holding at the HIP temperature, following furnace cooling, 2nd normalizing to refine austenite grains, and 2nd tempering. Microstructural evolutions during these various heat treatments should be focused, because they determine the performance of the blanket module. Especially, fine precipitates such as tantalum and vanadium carbides precipitated at high temperatures greatly affect the creep property, the material toughness, and irradiation resistances of RAF as the structural material. In this work, we have studied the stability of fine precipitates in the F82H-BA07 heat (8Cr-2W-V, Ta) during simulated heat treatments of the blanket fabrication.
Nakamichi, Masaru; Yonehara, Kazuo; Wakai, Daisuke
Fusion Engineering and Design, 86(9-11), p.2262 - 2264, 2011/10
Times Cited Count：28 Percentile：86.90(Nuclear Science & Technology)Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Ida, Mizuho; Yagi, Juro*; Suzuki, Akihiro*; Fukada, Satoshi*; Matsushita, Izuru*; Nakamura, Kazuyuki
Fusion Engineering and Design, 86(9-11), p.2437 - 2441, 2011/10
Times Cited Count：24 Percentile：83.60(Nuclear Science & Technology)Engineering Validation and Engineering Design Activities (EVEDA) for The International Fusion Materials Irradiation Facility (IFMIF) were started from July 2007 under an international agreement called ITER Broader Approach. As a major Japanese activity, EVEDA Li test loop (ELTL) to simulate hydraulic and impurity conditions of IFMIF has already designed and is under construction, in which feasibility of hydraulic stability of the liquid Li target, the purification systems of hot traps are major key issues to be validated in this loop. This paper focuses on the purification systems of the ELTL. Design of a cold trap and hot traps are discussed in this paper.
Coletti, A.*; Baulaigue, O.*; Cara, P.*; Coletti, R.*; Ferro, A.*; Gaio, E.*; Matsukawa, Makoto; Novello, L.*; Santinelli, M.*; Shimada, Katsuhiro; et al.
Fusion Engineering and Design, 86(6-8), p.1373 - 1376, 2011/10
Times Cited Count：24 Percentile：83.60(Nuclear Science & Technology)Watanabe, Kazuyoshi; Ida, Mizuho; Kondo, Hiroo; Nakamura, Kazuyuki; Wakai, Eiichi
Fusion Engineering and Design, 86(9-11), p.2482 - 2486, 2011/10
Times Cited Count：2 Percentile：17.23(Nuclear Science & Technology)The Engineering Validation and Engineering Design Activities (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) are in progress under the Broader Approach (BA) Agreement. As a part of this engineering design, we carried out thermo-structural analysis of the back plate in the IFMIF target. In this analysis, the target assembly of the integrated back plate option was modeled with the nuclear heating to simulate the IFMIF usual operation. The calculation parameters were thermal boundary conditions of a mechanical joint between the target assembly and the beam duct. The calculation results showed the influence of parameters on thermal stress was small. The maximum von Mises stresses occurred at the back plate center and those values, 204 - 218 MPa were lower than half of the yield strength of F82H (455 MPa). The maximum thermal deformations occurred at the same place and those values, about 0.3 mm will be important input parameter for the Li flow stability analysis.
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Sato, Satoshi; Seki, Yohji; Takase, Haruhiko
Fusion Engineering and Design, 86(9-11), p.2378 - 2381, 2011/10
Times Cited Count：12 Percentile：63.78(Nuclear Science & Technology)For DEMO reactor blanket design, a two-dimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed. In DOHEAT, the neutron flux is calculated by a 2-D transport code, DOT3.5, with the nuclear data library, FUSION-40, and the nuclear heating rate and the local TBR profile of blanket are calculated using the 2-D neutronics calculation code, APPLE-3. Use of the code has showed outstanding usefulness in the blanket design where detailed evaluation of neutron flux, nuclear heating rate, tritium breeding ratio (TBR) and the temperature of materials is required for various blanket concepts and trial-and-error-basis iteration is sometimes necessary. DOHEAT can replace the actual blanket structure by a more realistic model including cooling tubes, multipliers and breeders. A validation calculation indicates that DOHEAT provides reasonable results on the temperature profile.
Shimada, Katsuhiro; Baulaigue, O.*; Cara, P.*; Coletti, A.*; Coletti, R.*; Matsukawa, Makoto; Terakado, Tsunehisa; Yamauchi, Kunihito
Fusion Engineering and Design, 86(6-8), p.1427 - 1431, 2011/10
Times Cited Count：11 Percentile：60.99(Nuclear Science & Technology)Nakamura, Kazuyuki; Furukawa, Tomohiro; Hirakawa, Yasushi; Kanemura, Takuji; Kondo, Hiroo; Ida, Mizuho; Niitsuma, Shigeto; Otaka, Masahiko; Watanabe, Kazuyoshi; Horiike, Hiroshi*; et al.
Fusion Engineering and Design, 86(9-11), p.2491 - 2494, 2011/10
Times Cited Count：11 Percentile：60.99(Nuclear Science & Technology)In IFMIF/EVEDA, tasks for lithium target system are shared to 5 validation tasks (LF1-5) and a design task (LF6). The purpose of LF1 task is to construct and operate the EVEDA lithium test loop, and JAEA has a main responsibility to the performance of the Li test loop. LF2 is a task for the diagnostics of the Li test loop and IFMIF design. Basic research for the diagnostics equipment has been completed, and the construction for the Li test loop will be finished before March in 2011. LF4 is a task for the purification systems with nitrogen and hydrogen. Basic research for the purification equipment has been completed, and the construction of the nitrogen system for the Li test loop will be finished before March in 2011. LF5 is a task for the remote handling system with the target assembly. JAEA has an idea to use the laser beam for cutting and welding of the lip part of the flanges. LF6 is a task for the design of the IFMIF based on the validation experiments of LF1-5.
