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Journal Articles

The Current status of the world ITER test blanket module program

Konishi, Satoshi*; Enoeda, Mikio

Purazuma, Kaku Yugo Gakkai-Shi, 90(6), p.332 - 337, 2014/06

Test Blanket Module (TBM) program is to evaluate important functions of prototypical modules of DEMO breeding blankets in the real DT fusion plasma environment of ITER. Therefore, it is regarded as one of the most important milestones toward DEMO blanket. Japan is proposing a Water Cooled Ceramic Breeder (WCCB) TBM as the primary option of TBM program. Japan Atomic Energy Agency (JAEA) is performing the development of the WCCB blanket as the candidate breeding blanket of Japan, with a collaboration of universities and National Institute for Fusion Science (NIFS). Regarding the TBM development, the engineering R and Ds are ongoing, aiming at the demonstration of fabrication technology and structural integrity of the full size mockup of the WCCB TBM. Regarding the test blanket module fabrication technology development, the real scale back wall mockup was successfully fabricated. Also, the design activities are being performed to show the soundness under various loading conditions of electromagnetic force and thermo-mechanical loading. The evaluation of shutdown dose rate behind the TBM test port is also carried out as one of most important design requirement. Furthermore, key technologies toward DEMO blanket, such as, the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li$$_{2}$$TiO$$_{3}$$ pebble and BeTi pebble was performed.

Journal Articles

Effects of irradiation on mechanical properties of HIP-bonded reduced-activation ferritic/martensitic steel F82H first wall

Furuya, Kazuyuki; Wakai, Eiichi; Miyamoto, Kenji*; Akiba, Masato; Sugimoto, Masayoshi

Journal of Nuclear Materials, 367-370(1), p.494 - 499, 2007/08

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

A partial mock-up of a breeding blanket structure made of F82H steel has been successfully fabricated. In this study, microstructural observation and EDX analysis of the HIP interfaces were performed, and effects of irradiation on mechanical properties of the HIP-bonded region were also examined. Neutron irradiation was performed up to about 2 dpa at about 523 K. After the irradiation, tensile test was performed at temperatures of 295 and 523 K. The HIP interfaces possessed many precipitates, and enriched peak spectrum of chromium was detected from the precipitates. In addition, aspect of the spectrum was qualitatively equivalent to that of M$$_{23}$$C$$_{6}$$ in grain boundaries of F82H steel. In result, the HIP boundary has many M$$_{23}$$C$$_{6}$$ which were generally seen in grain boundaries of F82H steel. Rupture did not occur in the HIP interface. In result, it can be mentioned that bondability is maintained under the irradiation and testing conditions. The strength and elongation of the HIP-bonded region decreased somewhat in comparison with the results of an IEA standard steel.

Journal Articles

Neutron irradiation effect on mechanical properties of SS/SS HIP joint materials for ITER shielding blankets

Yamada, Hirokazu*; Sato, Satoshi; Mori, Kensuke*; Nagao, Yoshiharu; Takada, Fumiki; Kawamura, Hiroshi

Fusion Engineering and Design, 81(1-7), p.631 - 637, 2006/02

 Times Cited Count:1 Percentile:88.56(Nuclear Science & Technology)

This study estimated the neutron irradiation effect with 1.5 dpa on the mechanical properties of the SS/SS HIP joint materials jointed in the standard HIP joint condition. Results of this study showed that the HIP process in the standard HIP condition could make SS/SS HIP joint material of which tensile properties was equivalent to that of the SS base material. In addition, the effect of surface roughness at the HIP joint material on the mechanical properties of SS/SS HIP joint material was estimated.

Journal Articles

Evaluation of contact strength of Li$$_{2}$$TiO$$_{3}$$ pebbles with different diameters

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Tanaka, Satoru*

Fusion Engineering and Design, 81(8-14), p.1065 - 1069, 2006/02

 Times Cited Count:9 Percentile:44.48(Nuclear Science & Technology)

no abstracts in English

Journal Articles

A Design study for tritium recovery system from cooling water of a fusion power plant

Yamanishi, Toshihiko; Iwai, Yasunori; Kawamura, Yoshinori; Nishi, Masataka

Fusion Engineering and Design, 81(1-7), p.797 - 802, 2006/02

no abstracts in English

Journal Articles

Progress in the blanket neutronics experiments at JAERI/FNS

Sato, Satoshi; Verzilov, Y. M.; Ochiai, Kentaro; Nakao, Makoto*; Wada, Masayuki*; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori*; Nishitani, Takeo

Fusion Engineering and Design, 81(8-14), p.1183 - 1193, 2006/02

 Times Cited Count:14 Percentile:30.24(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Design study of fusion DEMO plant at JAERI

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Sato, Masayasu; Isono, Takaaki; Sakurai, Shinji; Nakamura, Hirofumi; Sato, Satoshi; Suzuki, Satoshi; Ando, Masami; et al.

Fusion Engineering and Design, 81(8-14), p.1151 - 1158, 2006/02

 Times Cited Count:107 Percentile:0.68(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Consideration on blanket structure for fusion DEMO plant at JAERI

Nishio, Satoshi; Omori, Junji*; Kuroda, Toshimasa*; Tobita, Kenji; Enoeda, Mikio; Tsuru, Daigo; Hirose, Takanori; Sato, Satoshi; Kawamura, Yoshinori; Nakamura, Hirofumi; et al.

Fusion Engineering and Design, 81(8-14), p.1271 - 1276, 2006/02

 Times Cited Count:13 Percentile:27.84(Nuclear Science & Technology)

The design guideline for the blanket is decided to meet the mission of the DEMO plant which is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. To moderate the technological extrapolation, the structural material of reduced activation ferritic steel (F82H), ceramic tritium breeder of Li$$_{2}$$TiO$$_{3}$$ and neutron multiplier of Be are introduced. To improve the economical aspect, the coolant material of the supercritical water with inlet/outlet temperatures of 280/510$$^{circ}$$C, coolant pressure of 25 MPa is chosen. Resultantly the thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirement (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.

Journal Articles

Effects of gelation and sintering conditions on granulation of Li$$_{2}$$TiO$$_{3}$$ pebbles from Li-Ti complex solution

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Casadio, S.*; Alvani, C.*

Fusion Engineering and Design, 75-79, p.877 - 880, 2005/11

 Times Cited Count:22 Percentile:15.63(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Behavior on Li$$_{2}$$TiO$$_{3}$$ under varied surface condition

Olivares, R.*; Oda, Takuji*; Oya, Yasuhisa*; Tanaka, Satoru*; Tsuchiya, Kunihiko

Fusion Engineering and Design, 75-79, p.765 - 768, 2005/11

 Times Cited Count:8 Percentile:45.03(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Selection of design solutions and fabrication methods and supporting R&D for procurement of ITER vessel and FW/blanket

Ioki, Kimihiro*; Elio, F.*; Maruyama, So; Morimoto, Masaaki*; Rozov, V.*; Tivey, R.*; Utin, Y.*

Fusion Engineering and Design, 74(1-4), p.185 - 190, 2005/11

 Times Cited Count:5 Percentile:59.91(Nuclear Science & Technology)

The ITER project has started preparation of Procurement Specification Documents for the vacuum vessel (VV). The design of the VV and FW/Blanket has progressed in many aspects, such as an double curvature pressing instead of facet shape welding for inner and outer shells in the upper and lower inboard regions to improve the fabrication and NDT process. The plasma facing surface of the FW has been defined to avoid protruding the leading edges, especially in the inboard area. Separate FW panels are supported with a central beam, and selection of a race-track shape cross-section for the central beam provides a more robust structure against halo current EM loads and also leads to a new cooling configuration in the shield block, where the pressure drop is significantly reduced to $$sim$$0.05 MPa. A UT R&D program is also going on to achieve acceptable S/N ratio for small-angle launching waves (20-30 deg.) to a weld. Hydraulic testing has been performed to demonstrate natural convection cooling in the transient condition.

Journal Articles

Control of particle size and density of Li$$_{2}$$TiO$$_{3}$$ pebbles fabricated by indirect wet processes

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Takayama, Tomoo*; Kato, Shigeru*

Journal of Nuclear Materials, 345(2-3), p.239 - 244, 2005/10

 Times Cited Count:34 Percentile:7.81(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Intelligible seminar on fusion reactors, 7; Optimum materials for the realization of fusion reactor

Hasegawa, Akira*; Tsuchiya, Kunihiko; Ishitsuka, Etsuo

Nippon Genshiryoku Gakkai-Shi, 47(8), p.536 - 544, 2005/08

no abstracts in English

JAEA Reports

Li depletion effects on Li$$_{2}$$TiO$$_{3}$$ reaction with H$$_{2}$$ in thermo-chemical environment relevant to breeding blanket for fusion power plants

Alvani, C.*; Casadio, S.*; Contini, V.*; Giorgi, R.*; Mancini, M. R.*; Tsuchiya, Kunihiko; Kawamura, Hiroshi

JAERI-Review 2005-024, 28 Pages, 2005/07

JAERI-Review-2005-024.pdf:5.73MB

no abstracts in English

Journal Articles

Interlinked test results for fusion fuel processing and blanket tritium recovery systems using cryogenic molecular sieve bed

Yamanishi, Toshihiko; Hayashi, Takumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Uzawa, Masayuki*; Nishi, Masataka

Fusion Science and Technology, 48(1), p.63 - 66, 2005/07

 Times Cited Count:6 Percentile:59.91(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Experimental studies on tungsten-armour impact on nuclear responses of solid breeding blanket

Sato, Satoshi; Nakao, Makoto*; Verzilov, Y. M.; Ochiai, Kentaro; Wada, Masayuki*; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori*; Nishitani, Takeo

Nuclear Fusion, 45(7), p.656 - 662, 2005/07

 Times Cited Count:8 Percentile:67.3(Physics, Fluids & Plasmas)

no abstracts in English

Journal Articles

Neutronics experiments using small partial mockups of the ITER test blanket module with a solid breeder

Sato, Satoshi; Verzilov, Y. M.; Nakao, Makoto*; Ochiai, Kentaro; Wada, Masayuki*; Nishitani, Takeo

Fusion Science and Technology, 47(4), p.1046 - 1051, 2005/05

 Times Cited Count:12 Percentile:31.6(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Low temperature tritium release experiment from lithium titanete breeder material

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nakamichi, Masaru*; Sagawa, Hisashi

JAERI-Tech 2005-013, 56 Pages, 2005/03

JAERI-Tech-2005-013.pdf:6.4MB

no abstracts in English

JAEA Reports

Achievements of element technology development for breeding blanket

Department of Fusion Engineering Research; Department of Materials Science

JAERI-Review 2005-012, 143 Pages, 2005/03

JAERI-Review-2005-012.pdf:11.74MB

no abstracts in English

Journal Articles

Fusion materials and hydrogen

Nagasaki, Takanori*; Yamaguchi, Kenji; Konishi, Satoshi*

Nippon Genshiryoku Gakkai-Shi, 46(11), p.770 - 779, 2004/11

no abstracts in English

302 (Records 1-20 displayed on this page)