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Sato, Satoshi; Iida, Hiromasa; Ochiai, Kentaro; Konno, Chikara; Nishitani, Takeo; Morota, Hidetsugu*; Nashif, H.*; Yamada, Masao*; Masuda, Fukuzo*; Tamamizu, Shigeyuki*; et al.
Nuclear Technology, 168(3), p.843 - 847, 2009/12
Times Cited Count：7 Percentile：43.22(Nuclear Science & Technology)It takes huge or unrealistic amounts of time to prepare accurate calculation inputs in shielding design for very large and complicated structure such as fusion reactors. For that reason, we have developed an automatic conversion system from three dimensional CAD drawing data into input data of the calculation geometry for a three dimensional Monte Carlo radiation transport calculation code MCNP, and applied it to an ITER benchmark model. This system consists of a void creation program (CrtVoid) for CAD drawing data and a conversion program (GEOMIT) from CAD drawing data to MCNP input data. CrtVoid creates void region data by subtracting solid region data from the whole region by Boolean operation. The void region data is very large and complicated geometry. The program divides the overall region to many small cubes, and the void region data can be created in each cube. GEOMIT generates surface data for MCNP data based on the CAD data with voids. These surface data are connected, and cell data for MCNP input data are generated. In generating cell data, additional surfaces are automatically created in the program, and undefined space and duplicate cells are removed. We applied this system to the ITER benchmark model. We successfully created void region data, and MCNP input data. We calculated neutron flux and nuclear heating. The calculation results agreed well with those with MCNP inputs generated from the same CAD data with other methods.
Shaaban, N.*; Masuda, Fukuzo*; Nasif, H.*; Yamada, Masao*; Sawamura, Hidenori*; Morota, Hidetsugu*; Sato, Satoshi; Iida, Hiromasa; Nishitani, Takeo
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 7 Pages, 2006/07
no abstracts in English
Ioki, Kimihiro*; Barabaschi, P.*; Barabash, V.*; Chiocchio, S.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Gervash, A.*; Ibbott, C.*; Jones, L.*; et al.
Nuclear Fusion, 43(4), p.268 - 273, 2003/04
Times Cited Count：21 Percentile：53.22(Physics, Fluids & Plasmas)Although the basic concept of the vacuum vessel (VV) and in-vessel components of the ITER design has stayed the same, there have been several detailed design improvements resulting from efforts to raise reliability, to improve maintainability and to save money. One of the most important achievements in the VV R&D has been demonstration of the necessary fabrication and assembly tolerances. Recently the deformation due to cutting of the port extension was measured and it was shown that the deformation is small and acceptable. Further development of advanced methods of cutting, welding and NDT on a thick plate have been continued in order to refine manufacturing and improve cost and technical performance. With regard to the related FW/blanket and divertor designs, the R&D has resulted in the development of suitable technologies. Prototypes of the FW panel, the blanket shield block and the divertor components have been successfully fabricated.
Ioki, Kimihiro*; Akiba, Masato; Cardella, A.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Lorenzetto, P.*; Miki, Nobuharu*; Osaki, Toshio*; Rozov, V.*; et al.
Fusion Engineering and Design, 61-62, p.399 - 405, 2002/11
Times Cited Count：11 Percentile：56.66(Nuclear Science & Technology)We report progress on the ITER-FEAT Blanket design and R&D during 2001-2002. Four major sub-components (FW, shield body, flexible support and electrical connection) have been highlighted. Regarding the FW, design on a separate FW panel concept has progressed, and heat load tests on a small-scale mock-up have been successfully performed with 0.7 MW/m
, 13000 cycles. Full-scale separate FW panels (dimensions: 0.9
0.250.07 m) have been fabricated by HIPing and brazing. Regarding the shield body, a radial flow cooling design has been developed, and full-scale partial mock-ups have been fabricated by using water-jet cutting. A separate FW panel was assembled with one the shield body mock-ups. Regarding the flexible support, mill-annealed Ti (easier fabricability) alloy has been selected, and the remote assembly has been considered in the design. In mechanical tests, the requires buckling strength and mechanical fatigue characteristics have been confirmed. Regarding the electrical connection, one-body structure design without welding joints has been developed. Mechanical fatigue tests in the 3 directions, have been carried out, and thermal fatigue tests and electrical tests in a solenoidal magnetic field have been performed. Feasibility of the design has been confirmed. Through progress on design and R&D of the blanket, cost reduction has been achieved, and feasibility of design and fabricability of the components have been confirmed.
Nishio, Satoshi; ; ; ; ; Koizumi, Koichi; Abe, Tetsuya; ; Tada, Eisuke
JAERI-M 91-089, 138 Pages, 1991/05
no abstracts in English
Kikuchi, Mitsuru; Seki, Yasushi; Oikawa, Akira; Ando, Toshinari; Ohara, Yoshihiro; Nishio, Satoshi; Seki, Masahiro; Takizuka, Tomonori; Tani, Keiji; Ozeki, Takahisa; et al.
Fusion Engineering and Design, 18, p.195 - 202, 1991/00
Times Cited Count：9 Percentile：68.72(Nuclear Science & Technology)no abstracts in English
Omori, Junji*; Kobayashi, Takeshi; Yamada, Masao*; Iida, Hiromasa; Horie, Tomoyoshi
Fusion Engineering and Design, 9, p.207 - 211, 1989/00
no abstracts in English
; Adachi, Junichi*; Iida, Hiromasa; ; ; ;
IAEA-TECDOC-495, p.51 - 62, 1989/00
no abstracts in English
; ; ; Iida, Hiromasa
JAERI-M 88-081, 29 Pages, 1988/05
no abstracts in English
Mizoguchi, Tadanori*; Iida, Hiromasa; Sugihara, Masayoshi; Fujisawa, Noboru; ; ; ; Nakajima, Kunihiko*; Nishio, Satoshi; ; et al.
JAERI-M 88-062, 77 Pages, 1988/03
no abstracts in English
Iida, Hiromasa; ; ; Adachi, Junichi*; ; ; Hamajima, Takataro*; Hatayama, Akiyoshi*; ; ; et al.
JAERI-M 88-011, 261 Pages, 1988/02
no abstracts in English
Sugihara, Masayoshi; ; Nishio, Satoshi; ; Yamamoto, Shin;
JAERI-M 87-216, 24 Pages, 1988/01
no abstracts in English
; ; Mizoguchi, Tadanori*; ; ; ; Watanabe, Takashi*; Mori, Seiji*; Adachi, Junichi*; ; et al.
JAERI-M 87-139, 355 Pages, 1987/09
no abstracts in English
; ; Mizoguchi, Tadanori*; ; ; ; Watanabe, Takashi*; Mori, Seiji*; Adachi, Junichi*; ; et al.
JAERI-M 87-138, 155 Pages, 1987/09
no abstracts in English
; ; ; ; Nakajima, Kunihiko*; Sugihara, Masayoshi; Yamamoto, Shin; Iida, Hiromasa; Fujisawa, Noboru; Mizoguchi, Tadanori*; et al.
JAERI-M 87-137, 72 Pages, 1987/09
no abstracts in English
Mizoguchi, Tadanori*; Sugihara, Masayoshi; Shinya, K.*; ; ; Nakajima, Kunihiko*; ; ; Fujisawa, Noboru; Yamamoto, Shin; et al.
JAERI-M 87-120, 49 Pages, 1987/08
no abstracts in English
; ; ; ; Fujisawa, Noboru
JAERI-M 87-112, 38 Pages, 1987/08
no abstracts in English
Iida, Hiromasa; ; Yamada, Masao*; Suzuki, Tatsushi*; Honda, Tsutomu*; Omura, Hiroshi*; Ito, Shinichi*
JAERI-M 86-149, 314 Pages, 1986/11
This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied sith careful zoning concept by the radiation level. Structual integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the shole FER plant system is developed.
; ; ; ; ; Kondo, I.; ; ; ; Tsunematsu, Toshihide; et al.
JAERI-M 85-077, 203 Pages, 1985/07
no abstracts in English
Tobita, Kenji; Saito, Ai*; Nishio, Satoshi; Enoeda, Mikio; Tanigawa, Hisashi; Sato, Satoshi; Tsuru, Daigo; Hirose, Takanori; Seki, Yohji; Yamada, Masao*
no journal, ,
Self-sufficiency of tritium fuel is an indispensable requirement in a fusion reactor. In this sense, the blanket design ensuring the self-sufficiency of fuel should be considered with importance and higher priority in fusion reactor design. Generally speaking, however, it is difficult to envisage a self-consistent blanket design satisfying various requirements such as robustness to an enormous electromagnetic forces and excellent heat removal performance other than fuel self-sufficiency with matured candidate materials. This paper presents a concept study on torus configuration with water-cooled solid breeder blanket satisfying these functions. The result indicates that (1) the conducting shell should be arranged at the position of r
/a =1.34 or farther, and (2) when blanket is installed behind the divertor, the increment of tritium breeding ratio is as low as 0.01-0.02. In addition, promising blanket concepts in the presently studied DEMO reactor are proposed.
