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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：24.26(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：93.85(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：46.21(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：57.67(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：91.52(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.
Hoshino, Tsuyoshi; Terai, Takayuki*
Fusion Engineering and Design, 86(9-11), p.2168 - 2171, 2011/10
Times Cited Count：62 Percentile：96.81(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.
Nakamichi, Masaru; Yonehara, Kazuo; Wakai, Daisuke
Fusion Engineering and Design, 86(9-11), p.2262 - 2264, 2011/10
Times Cited Count：28 Percentile：86.88(Nuclear Science & Technology)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：41.46(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.
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：17.22(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.
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：60.87(Nuclear Science & Technology)Nozawa, Takashi; Ogiwara, Hiroyuki*; Kannari, Jun*; Kishimoto, Hirotatsu*; Tanigawa, Hiroyasu
Fusion Engineering and Design, 86(9-11), p.2512 - 2516, 2011/10
Times Cited Count：14 Percentile：69.10(Nuclear Science & Technology)A hot isostatic press (HIP) process is a key technology to fabricate a first wall of the blanket system utilizing a reduced-activation ferritic/martensitic (RAFM) steel such as F82H. A primary objective of this study is to characterize interfacial properties of HIPed F82H joints by torsion to identify the feasibility of this test method. It is apparent that the absorption energies of the HIP joints varied by the processing conditions, although the maximum shear strength was not much different. According to the fracture surfaces, it is indicated that the reduction of the absorption energy was due to the oxide formed on the interface of the HIP joint and this was consistent with the results of charpy impact tests. In conclusion, the torsion test method enables to precisely evaluate the shear properties of the HIPed joint interface and becomes one of promising powerful techniques for inspection of the HIP joints.
Hirose, Takanori; Tanigawa, Hisashi; Yoshikawa, Akira; Seki, Yohji; Tsuru, Daigo; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato
Fusion Engineering and Design, 86(9-11), p.2265 - 2268, 2011/10
Times Cited Count：5 Percentile：36.25(Nuclear Science & Technology)As one of the most important fabrication technologies of the WCCB TBM, Hot Isostatic Pressing (HIP) joining technology was selected to fabricate the first wall with built-in cooling channel structure made of reduced activation martensitic/ferritic steel, F82H. By using developed HIP technology, a real scale TBM first wall mockup was successfully fabricated. High heat flux test of the fabricated mockup showed the feasibility to with the equivalent conditions of the WCCB TBM operation. The breeder pebble box was successfully fabricated with thin wall cooling pipes and thin plate sleds by Laser welding. With respect to the side walls with built in cooling channels were also fabricated using drilling technology. Assembling of the first wall and side walls is one of the critical fabrication processes of the fabrication of the TBM structure. By using a F82H first wall mockup and side wall mockups, assembling process was demonstrated successfully by Electron Beam welding.
-ray and neutron area monitoring system for the IFMIF/EVEDA accelerator buildingTakahashi, Hiroki; Maebara, Sunao; Kojima, Toshiyuki; Kubo, Takashi; Sakaki, Hironao; Takeuchi, Hiroshi; Shidara, Hiroyuki; Hirabayashi, Keiichi*; Hidaka, Kosuke*; Shigyo, Nobuhiro*; et al.
Fusion Engineering and Design, 86(9-11), p.2795 - 2798, 2011/10
Times Cited Count：2 Percentile：17.22(Nuclear Science & Technology)In the IFMIF/EVEDA accelerator, the engineering validation up to 9 MeV by employing the deuteron beam of 125 mA are planning at the BA site in Rokkasho, Aomori, Japan, the personnel protection system (PPS) is indispensable. The PPS inhibit the beam by receiving the interlock signal from the -ray and neutron monitoring system. The -ray and neutron detection level which is planned to be adopted are "80 keV to 1.5 MeV (-ray)" and "0.025 eV to 15 MeV (neutron)". For the present shielding design, it is absolutely imperative for the safety review to validate the shielding ability which makes detection level lower than these -ray and neutron detector. For this purpose, the energy reduction of neutron and photon for water and concrete is evaluated by PHITS code. From the calculating results, it is found that the photon energy range extended to 10 MeV by water and concrete shielding material only, an additional shielding to decrease the photon energy of less than 1.5 MeV is indispensable.
Kizu, Kaname; Kashiwa, Yoshitoshi; Murakami, Haruyuki; Obana, Tetsuhiro*; Takahata, Kazuya*; Tsuchiya, Katsuhiko; Yoshida, Kiyoshi; Hamaguchi, Shinji*; Matsui, Kunihiro; Nakamura, Kazuya*; et al.
Fusion Engineering and Design, 86(6-8), p.1432 - 1435, 2011/10
Times Cited Count：8 Percentile：50.52(Nuclear Science & Technology)In JT-60SA, central solenoid (CS) and plasma equilibrium field (EF) coils are procured by Japan. EF coil conductors are NbTi cable-in-conduit (CIC) conductor. Delivered superconducting cables and jackets are fabricated into CIC conductors at the jacketing facility with the length of 680 m constructed in the Naka site of JAEA. The production of superconductors with 444 m in length for actual EF coils was started from March 2010. The measurements of superconducting performance like current sharing temperature (Tcs) were conducted prior to the mass production. The measured Tcs was agreed with the expectation values from strand values indicating that no degradation was happened by production process.
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：10.05(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.
Takahashi, Koji; Kajiwara, Ken; Okazaki, Yukio*; Oda, Yasuhisa; Sakamoto, Keishi; Omori, Toshimichi*; Henderson, M.*
Fusion Engineering and Design, 86(6-8), p.982 - 986, 2011/10
Times Cited Count：7 Percentile：46.21(Nuclear Science & Technology)In order to optimize the physics performance of ITER, the millimeter (mm) wave design of the equatorial electron cyclotron launcher is modified so that one-third of the total mm-wave injection power can be flipped to drive plasma current in the counter direction. The design change to perform the poloidal beam tilting of 5 at top and bottom beam row is conducted so that more efficient central power deposition is achievable. The opening shape in the blanket shield module (BSM) structure is also optimized to minimize the degradation of the mm-wave transmission efficiency. The degradation is only 0.3%. The modularization design of the steering mirror components and the internal shield, the application of a rail structure to perform the alignment of the steering mirror installation and the space allocation behind the BSM structure are implemented to increase the reliability, the manufacturability and the maintenability of the launcher design.
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：74.97(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.
Iwai, Yasunori; Sato, Katsumi; Yamanishi, Toshihiko
Fusion Engineering and Design, 86(9-11), p.2164 - 2167, 2011/10
Times Cited Count：19 Percentile：77.81(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.
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：63.66(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.
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：83.56(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.
