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Teshigawara, Makoto; Tsuchikawa, Yusuke*; Ichikawa, Go*; Takata, Shinichi; Mishima, Kenji*; Harada, Masahide; Oi, Motoki; Kawamura, Yukihiko*; Kai, Tetsuya; Kawamura, Seiko; et al.
Nuclear Instruments and Methods in Physics Research A, 929, p.113 - 120, 2019/06
Times Cited Count：4 Percentile：20.8(Instruments & Instrumentation)A nano-diamond is an attractive neutron reflection material below cold neutron energy. The total neutron cross section of a nano-diamond was derived from a neutron transmission measurement over the neutron energy range of 0.2 meV to 100 meV because total neutron cross section data were not available. The total cross section of a nano-diamond with particle size of approximately 5 nm increased with a decrease in neutron energy to 0.2 meV. It was approximately two orders of magnitude larger than that of graphite at 0.2 meV. The contribution of inelastic scattering to the total cross section was to be shown negligible small at neutron energies of 1.2, 1.5, 1.9, 2.6, and 5.9 meV in the inelastic neutron scattering measurement. Moreover, small-angle neutron scattering measurements of the nano-diamond showed a large scattering cross section in the forward direction for low neutron energies.
Nakajima, Kenji; Kawamura, Seiko; Kikuchi, Tatsuya*; Kofu, Maiko; Kawakita, Yukinobu; Inamura, Yasuhiro; Kambara, Wataru*; Aoyama, Kazuhiro*; Wakai, Daisuke*; Harada, Masahide; et al.
Journal of Physics; Conference Series, 1021(1), p.012031_1 - 012031_5, 2018/06
Times Cited Count：3 Percentile：4.98Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
Journal of Plasma and Fusion Research SERIES, Vol.11, p.69 - 72, 2015/03
Water-cooled blanket is an attractive concept for its compactness and its compatibility with the conventional technologies for PWR. For blanket application, the structural material is required to be as thin as possible for tritium breeding. On the other hand, it is also required the pressure tightness to withstand 15 MPa of internal pressure. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. The effects of water flow and DO in the test water on corrosion properties were investigated using rotating disk specimen in autoclave. In summary, the weight loss by flowing was occurred except for test with DO 8 ppm, and it was more pronounced at lower DO concentration. Since Fe
O
was observed on the specimen of small weight change, and the iron-poor layer thickness increased with decreasing the specimen weight, it seemed that the formation of FeO was effective for the suppression of weight loss.
Kanai, Akihiko*; Kasada, Ryuta*; Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio; Konishi, Satoshi*
Journal of Nuclear Materials, 455(1-3), p.431 - 435, 2014/12
Times Cited Count：0 Percentile：100(Materials Science, Multidisciplinary)Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.
Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10
Times Cited Count：17 Percentile：12.12(Nuclear Science & Technology)The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.
Nakata, Toshiya; Komazaki, Shinichi*; Nakajima, Motoki*; Kono, Yutaka*; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Koyama, Akira*
Nippon Kinzoku Gakkai-Shi, 70(8), p.642 - 645, 2006/08
Times Cited Count：4 Percentile：63.48(Metallurgy & Metallurgical Engineering)no abstracts in English
Sb
Kogi, Masafumi*; Iwasa, Kazuaki*; Nakajima, Motoki*; Metoki, Naoto; Araki, Shingo; Bernhoeft, N.*; Mignot, J.-M.*; Gukasov, A.*; Sato, Hideyuki*; Aoki, Yuji*; et al.
Journal of the Physical Society of Japan, 72(5), p.1002 - 1005, 2003/05
Times Cited Count：192 Percentile：2.47(Physics, Multidisciplinary)no abstracts in English
Nakajima, Motoki; Kanai, Akihiko*; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
no journal, ,
no abstracts in English
Kanai, Akihiko*; Kasada, Ryuta*; Konishi, Satoshi*; Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
no journal, ,
no abstracts in English
Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Enoeda, Mikio
no journal, ,
no abstracts in English
Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
no journal, ,
no abstracts in English
Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Enoeda, Mikio
no journal, ,
no abstracts in English
Hirose, Takanori; Sakasegawa, Hideo; Nakajima, Motoki; Tanigawa, Hiroyasu
no journal, ,
Weld joints of a reduced activation ferritic/martensitic steel, F82H were prepared using Tungsten-Inert-Gas (TIG) and Electron Beam (EB) welding. Physical properties and mechanical properties of these weld joints were investigated in this work. Moreover, effects of Post Weld Heat Treatment (PWHT) was also investigated. After PWHT at 720 
C, most of physical properties of weld metal were very similar to those of F82H base metal. Coefficient of thermal expansion and thermal diffusivity of weld metal demonstrated 10% of degradation compared to the base metal. Although weld metal and heat affected zone heated above transformation temperature demonstrated hardening and embrittlement, PWHT above 750 C successfully moderated the hardening without softening in the base metal.
Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Kawamura, Yoshinori
no journal, ,
For blanket application, the structural material is required to be as thin as possible for tritium breeding. On the other hand, the pressure tightness is required to withstand 15 MPa of internal pressure. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. This work reports results of corrosion tests on a reduced activation ferritic/martensitic steel, F82H as a structural material for the blanket. Moreover, the effects of water flow and dissolved oxygen (DO) on corrosion properties were investigated using rotating disk specimen in autoclave. In previous study, it was reported that the weight loss by water flow becomes significant with lowering DO concentration. Based on the XRD and EPMA results, it seemed that the formation of FeO was effective for the suppression of weight loss. In this study, we discussed the relationship between stable oxide film and corrosion.
Nakajima, Motoki
no journal, ,
Reduced activation ferritic/martensitic steel, F82H is the most promising candidate structural material for fusion blanket. Water-cooled ceramic breeder blanket is an attractive concept for its compactness and its compatibility with the technologies in conventional pressurized water reactor (PWR). For blanket application, the structural material is required to be as thin as possible for tritium breeding. On the other hand, it is also required the pressure tightness which could withstand internal pressure of 15 MPa. Therefore it is necessary to determine corrosion allowance for cooling tubes precisely. This work reports corrosion properties of F82H in high temperature pressurized water. Moreover, the effects of water flow on corrosion properties were investigated using rotating disk specimen in autoclave.
Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Kawamura, Yoshinori
no journal, ,
As the primary candidate of ITER-TBM of Japan, development of WCCB TBM is being performed. In this blanket, the high temperature pressurized water was used as the coolant. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. Additionally, it is also required that understanding of corrosion properties of temperature ranging from 543K to 593K, because inlet and outlet temperatures of TBM are 543K and 593K, respectively. In this study, the corrosion test was performed at temperature ranging from 543K to 593K in high temperature water. Based on the results of corrosion test, the corrosion amount was increased with increasing the test temperature. The corrosion amount of outlet water was two times larger than that of inlet water.
Kawamura, Yoshinori; Hirose, Takanori; Tanigawa, Hisashi; Nakajima, Motoki; Gwon, H.; Miyata, Satoru; Sato, Satoshi
no journal, ,
Japan Domestic Agency (JADA) participates to the Test Blanket Module (TBM) test program of the ITER project as the lead party of Water Cooled Ceramic Breeder (WCCB) - TBM. We have signed TBM arrangement (TBMA) in Nov. 2014, and conceptual design has been reviewed in Feb. 2015. We have received 3 Cat-1 chits. Solution of Cat-1 chit is necessary to advance to the next design phase (preliminary design). Other intrinsic topic is the study on flow-accelerated corrosion of F82H. F82H is one of Reduced Activation Ferritic/Martensitic steels (RAFM) developed in Japan. In the piping of water cooling system, both F82H and stainless steel are used. As for fission reactor, there are many reports about corrosion of stainless steel. However, the report about F82H, especially flow-accelerated corrosion, is few. So, in this report, we will mainly show the situation for chit resolution, and the progress of the study on flow-accelerated corrosion of F82H.
Gwon, H.; Tanigawa, Hisashi; Nakajima, Motoki; Hirose, Takanori; Kawamura, Yoshinori
no journal, ,
It is concerned that the temperature of the blanket would increase excessively due to the decay heat even after the plasma shutdown. In present study, we focuses the relationship between the blanket structures and the passive cooling performance and considers how to effectively mitigate the excessive temperature rising due to the decay heat. The arrangement of both pebble beds and the ribs was changed as a way to relieve the temperature rising. The thermal response characteristics of the modified blanket models were evaluated by using a two-dimensional nuclear-thermal-coupled analysis code, DOHEAT3. The TBR of each model was also evaluated. Based on the results, the useful design policies for blanket were proposed from the viewpoint of the decay heat removal.
Nakajima, Motoki; Hirose, Takanori; Gwon, H.; Tanigawa, Hisashi; Kawamura, Yoshinori
no journal, ,
As the primary candidate of ITER-TBM of Japan, development of WCCB TBM is being performed. In this blanket, the high temperature pressurized water was used as the coolant. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. Additionally, it is also required that understanding of corrosion properties of temperature ranging from 543K to 593K, because inlet and outlet temperatures of TBM are 543K and 593K, respectively. In this study, the corrosion test was performed at temperature ranging from 543K to 593K in deaerated (20ppb DO) and oxygen-saturated (8ppm DO) high temperature water.
Tanigawa, Hisashi; Gwon, H.; Hirose, Takanori; Nakajima, Motoki; Kawamura, Yoshinori
no journal, ,
Blanket in fusion nuclear reactor has three functions such as neutron shielding, heat recovery and tritium breeding. Blankets with ceramic breeder materials has a box structure made of reduced activation ferritic/martensitic steel, and pebbles of breeder and neutron multiplier materials are packed into the box. Heat and neutron loads on the blanket are cooled by high temperature and pressure coolant such as water and helium. In this study, strength and pressure integrity are assessed for the water-cooled ceramic breeder blanket developed in JAEA. Existing design standards for pressure equipment are used in the analysis and their applicability to the fusion blanket is discussed.
