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Journal Articles

Development of small specimen test techniques for the IFMIF test cell

Wakai, Eiichi; Kim, B. J.; Nozawa, Takashi; Kikuchi, Takayuki; Hirano, Michiko*; Kimura, Akihiko*; Kasada, Ryuta*; Yokomine, Takehiko*; Yoshida, Takahide*; Nogami, Shuhei*; et al.

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 6 Pages, 2013/03

Journal Articles

Effects of lithium burn-up on TBR in DEMO reactor SlimCS

Sato, Satoshi; Nishitani, Takeo; Konno, Chikara

Fusion Engineering and Design, 87(5-6), p.680 - 683, 2012/08

 Times Cited Count:8 Percentile:52.86(Nuclear Science & Technology)

Lithium in a breeding blanket is burned up through neutron nuclear reactions in fusion DEMO reactors. For the SlimCS blanket design, the TBRs have been calculated taking into account the lithium burn-ups by one dimensional Sn radiation transport calculation code ANISN. Although the maximum value of the $$^{6}$$Li burn-up amounts to 79% after 10-years continuous operation, the total TBR in the blanket decrease to around 96% of the initial value. It is expected that the reduction of the TBR due to the lithium burn-up is not so large.

Journal Articles

Overview of materials research and IFMIF-EVEDA under the Broader Approach framework

Nishitani, Takeo; Tanigawa, Hiroyasu; Yamanishi, Toshihiko; Clement Lorenzo, S.*; Baluc, N.*; Hayashi, Kimio; Nakajima, Noriyoshi*; Kimura, Haruyuki; Sugimoto, Masayoshi; Heidinger, R.*; et al.

Fusion Science and Technology, 62(1), p.210 - 218, 2012/07

 Times Cited Count:3 Percentile:25.16(Nuclear Science & Technology)

Recent progress in the material related researches and the IFMIF/EVEDA project, which are carried out under the Broader Approach (BA) framework, is reported. In the International Fusion Energy Research Center (IFERC) project of BA, the R&D building was completed March 2010 at the Rokkasho BA site. R&Ds on reduced activation ferritic/ martensitic (RAFM) steels as structural material, SiC/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology relevant to the DEMO operational condition are progressed in Japan and EU. In the IFMIF/EVEDA project, the fabrication of the injector for the IFMIF prototype accelerator was completed at the CEA Saclay, and the first proton beam was obtained in May, 2011. The IFMIF lithium target test loop was completed in March 2011, and a lithium flow of 5 m/s was obtained.

Journal Articles

Japanese contribution to the DEMO-R&D program under the Broader Approach activities

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:49.27(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.

Journal Articles

IFMIF specifications from the users point of view

Garin, P.*; Diegele, E.*; Heidinger, R.*; Ibarra, A.*; Jitsukawa, Shiro; Kimura, Haruyuki; M$"o$slang, A.*; Muroga, Takeo*; Nishitani, Takeo; Poitevin, Y.*; et al.

Fusion Engineering and Design, 86(6-8), p.611 - 614, 2011/10

 Times Cited Count:26 Percentile:87.46(Nuclear Science & Technology)

This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the Users requirements and top level specifications for the Facility. Special attention is given to the different roadmaps of fusion path way towards power plants, of materials R&D and of facilities and their interactions. The materials development and validation activities on structural materials, blanket functional materials and non-metallic materials are analyzed and specific objectives and requirements to be implemented in IFMIF are proposed. Emphasis is made in additional potential validation activities that can be developed in IFMIF for ITER TBM qualification as well as for DEMO-oriented mock-up testing.

Journal Articles

Recent progress in blanket materials development in the Broader Approach Activities

Nishitani, Takeo; Tanigawa, Hiroyasu; Nozawa, Takashi; Jitsukawa, Shiro; Nakamichi, Masaru; Hoshino, Tsuyoshi; Yamanishi, Toshihiko; Baluc, N.*; M$"o$slang, A.*; Lindou, R.*; et al.

Journal of Nuclear Materials, 417(1-3), p.1331 - 1335, 2011/10

 Times Cited Count:14 Percentile:72.3(Materials Science, Multidisciplinary)

As a part of the Broader Approach (BA) activities, the research and development on blanket related materials and tritium technology have been initiated toward DEMO by Japan and EU. Recently, those five R&D items have progressed substantially in Japan and EU. As a preparatory work aiming at the RAFM steel muss-production development, a 5-ton heat of RAFM steel (F82H) was procured with the Electro Slag Re-melting as a secondary melting. The result of the double notch tensile test method for the NITE-SiC$$_{f}$$/SiC specimen indicated notch insensitivity and very minor size effect on proportional limit tensile stress and fracture strength. For the fabrication technology development of beryllide neutron multiplayer pebbles, Be- Ti inter-metallic pebbles have been sintered directly from the mixed powder of Be and Ti in Japan.

Journal Articles

Measurement of charged-particle emission double-differential cross section of fluorine for 14.2 MeV neutrons

Kondo, Keitaro; Murata, Isao*; Ochiai, Kentaro; Kubota, Naoyoshi*; Miyamaru, Hiroyuki*; Konno, Chikara; Nishitani, Takeo

Journal of Nuclear Science and Technology, 48(8), p.1146 - 1157, 2011/08

 Times Cited Count:3 Percentile:26.18(Nuclear Science & Technology)

We carried out a detailed measurement of the double-differential cross sections of fluorine for emitted protons, deuterons, tritons and $$alpha$$-particles with 14.2 MeV neutron incidence. An improved charged-particle spectrometer with a pencil DT-neutron beam furnished at the FNS facility of Japan Atomic Energy Agency enabled us to obtain precise data with a fine energy resolution in wide energy range and angular range from 15 to 150 $$^{circ}$$. The obtained data were compared with the nuclear data evaluated in JENDL-3.3 and ENDF/B-VII.0. As a result, large differences in the energy and angular distributions of emitted particles and the charged-particle production cross sections were found between the measured and evaluated data. Angular-differential cross sections for several discrete peaks corresponding to excited states of residual nuclei were extracted to discuss the reaction mechanism of charged-particle emission. The obtained data suggest that the charged-particle emission reaction of fluorine has a quite complicated mechanism where both the direct reaction process and the pre-equilibrium process contribute. The present experiment is the first simultaneous measurement of the four different kinds of charged particles and would provide useful data to confirm previous experimental data as well as to establish a nuclear reaction model of fluorine.

Journal Articles

Effect of thermal neutrons on fusion power measurement using the microfission chamber in ITER

Ishikawa, Masao; Kondoh, Takashi; Nishitani, Takeo; Kusama, Yoshinori

Fusion Engineering and Design, 86(4-5), p.417 - 420, 2011/06

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

A Microfission Chamber (MFC) provides time-resolved measurements of global neutron source strength in ITER. Measurements of the neutron source strength could be affected by cooling water in branch pipes installed near the MFC. The effect of the branch pipes upon the MFC is assessed through neutron transport calculation. Results indicate a significant increase in the MFC response rate due to the branch pipe. The increase in the MFC response is caused by the slowing down of the neutrons due to the cooling water in the branch pipes. One possible solution to reduce the effect is to cover the MFC with a material that absorbs thermal neutrons such as cadmium. The ways in which the absorbent material may affect MFC response is analyzed through neutron transport calculation. Results indicate that the increase in the MFC response can be reduced to $$<$$ 10% through cadmium coating.

Journal Articles

Parameters to determine the reactor performance

Nishitani, Takeo; Tobita, Kenji

Purazuma, Kaku Yugo Gakkai-Shi, 87(Suppl.), p.62 - 68, 2011/02

According to the reactor feasibility, minimum requirements, lifetime of each component and availability of the reactor are discussed. About the tritium breeding ratio which is a key factor for the fuel reproduction, coverage of the blanket is discussed not only by geometric coverage but also neutronics effects. As the lifetime of each component, radiation damage of the superconducting magnet, erosion of the armor material, radiation damage of the blanket structural material and burn-up of the tritium breeding material are discussed.

Journal Articles

Progress of IFERC project in the Broader Approach activities

Araki, Masanori; Sakamoto, Yoshiteru; Hayashi, Kimio; Nishitani, Takeo; Ouchi, Rei*

Fusion Engineering and Design, 85(10-12), p.2196 - 2202, 2010/12

 Times Cited Count:10 Percentile:56.44(Nuclear Science & Technology)

For contributing to the ITER project and promoting a possible early realization of DEMO, the IFERC project shall perform the activities on (1) DEMO Design and R&D Coordination, (2) Computational Simulation Centre, and (3) ITER Remote Experimentation Centre in the framework under the BA agreement. The DEMO design activity aims at establishing a common basis for DEMO design, including design features of DEMO, a possible common concept of DEMO, a roadmap for DEMO, and so on. Based on the common interest toward DEMO, the DEMO R&D activities have been planned and carried out for the following five areas which are relevant to blanket development: (1) SiC/SiC composites, (2) tritium technology, (3) materials engineering for DEMO blanket, (4) advanced neutron multiplier for DEMO blanket, and (5) advanced tritium breeders for DEMO blanket. In the activity of the Computational Simulation Centre, the objective is to provide and exploit a supercomputer for large scale simulation activities to analyse experimental data on fusion plasmas, prepare scenarios for ITER operation, predict the performance of the ITER facilities and contribute to the DEMO design. At the initial phase, high-level benchmark codes in the fusion research field have been selected through Special Working Group.

Journal Articles

Development of in-vessel components of the microfission chamber for ITER

Ishikawa, Masao; Kondoh, Takashi; Okawa, Kiyofumi*; Fujita, Kyoichi*; Yamauchi, Michinori*; Hayakawa, Atsuro*; Nishitani, Takeo; Kusama, Yoshinori

Review of Scientific Instruments, 81(10), p.10D308_1 - 10D308_3, 2010/10

 Times Cited Count:2 Percentile:13.34(Instruments & Instrumentation)

Microfission chambers (MFCs) will provide total neutron source strength in ITER. The MFC is a pencil-sized gas counter containing the fissile material, $$^{235}$$U. The MFCs will be installed behind blanket modules in the vacuum vessel (VV). Double coaxial mineral insulated (MI) cables will carry signals from the MFCs to the upper port. Though the MI cables will be installed at a factory of the vacuum vessel or ITER assembly hall, detectors with $$^{235}$$U will be installed to the vacuum vessel at the tokamak pit. Then, the MI cable should be connected in the vacuum vessel. In this work, the connection of the MI cable with the MFC was conceptually designed. The MI cable should be also installed with small curvature radius (R) of 100 $$sim$$ 200 mm to avoid the VV structure and other diagnostics. So, the vending test of the MI cable was conducted. As a result, damages leak, electrical disconnection and the change in insulation resistance have not been observed at R = 100 mm.

JAEA Reports

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08


This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m$$^{2}$$. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Journal Articles

Neutron transport analysis for in-vessel diagnostics in ITER

Ishikawa, Masao; Kondoh, Takashi; Nishitani, Takeo; Kawano, Yasunori; Kusama, Yoshinori

Journal of Plasma and Fusion Research SERIES, Vol.9, p.43 - 47, 2010/08

Neutron transport analysis is very important for design and optimization of diagnostics in ITER. Especially, in-vessel diagnostics are exposed to strong neutron and $$gamma$$ radiation and then it could lead to damage and temperature increase due to nuclear heating of the components of those diagnostics. High dose rate due to strong radiation also makes those maintenances difficult. Therefore, evaluation of neutron/$$gamma$$ flux, spectrum and nuclear heating at the location of the diagnostics with neutron transport analysis are essential to design a neutron radiation shield system and/or a cooling system. In this paper, results of neutron transport analysis applied to in-vessel components of the microfission chamber (MFC) and the poloidal polarimeter, which are developed by Japan Atomic Energy Agency, are presented.

Journal Articles

International Fusion Energy Research Centre

Araki, Masanori; Hayashi, Kimio; Tobita, Kenji; Nishitani, Takeo; Tanigawa, Hiroyasu; Nozawa, Takashi; Yamanishi, Toshihiko; Nakamichi, Masaru; Hoshino, Tsuyoshi; Ozeki, Takahisa; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 86(4), p.231 - 239, 2010/04

The Broader Approach Activities, which support the ITER Project and implement activities to aim early realization of fusion energy, is an EU-Japan collaborative project to carry out various kinds of researches and developments during the period of the ITER construction phase. In this special topic, achievements and prospects of the projects on the International Fusion Energy Research Centre (IFERC) is described.

Journal Articles

Development of CAD-to-MCNP model conversion system and its application to ITER

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:45.49(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.

Journal Articles

Development of the microfission chamber for fusion power diagnostics on ITER

Ishikawa, Masao; Kondoh, Takashi; Nishitani, Takeo; Kusama, Yoshinori

Journal of Plasma and Fusion Research SERIES, Vol.8, p.334 - 337, 2009/09

Microfission chambers (MFCs) are one of the most important diagnostics to measure total neutron source strength in ITER. The MFCs will be installed behind blanket modules upper outboard and lower outboard in the vacuum vessel. Double coaxial mineral insulated (MI) cables as signal cables are also installed form the MFCs to the upper port. It is very difficult to install the MI cables together with the MFC because of the security regulation. In this design work, a new type of MFC, which can be separable from the MI cable, has been designed. On the other hand, steaming neutrons along the gap between two blanket modules can affect the absolute measurement of total neutron source strength. The effects of streaming neutrons tat the installation position are investigated by a neutron Monte Carlo calculation using MCNP version 5 code. The result suggests that the effect of streaming neutrons should be taken into account if the MFCs are installed at the distance less than 20 cm from the gap.

Journal Articles

Status of advanced neutron multiplier development for DEMO in the Broader Approach activities in Japan

Nakamichi, Masaru; Yonehara, Kazuo; Nishitani, Takeo

Proceedings of 9th IEA International Workshop on Beryllium Technology (BeWS-9), p.16 - 20, 2009/09

Journal Articles

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

 Times Cited Count:135 Percentile:97.7(Physics, Fluids & Plasmas)

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m$$^{2}$$ or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).

Journal Articles

Direct neutron spectrum measurement to validate $$^{rm nat}$$Zr(n,2n) reaction cross-section at 14 MeV

Murata, Isao*; Shiken, Kimiaki*; Kondo, Keitaro; Matsunaka, Masayuki*; Ota, Masayuki*; Miyamaru, Hiroyuki*; Ochiai, Kentaro; Konno, Chikara; Nishitani, Takeo

Fusion Engineering and Design, 84(7-11), p.1376 - 1379, 2009/06

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Lithium zirconate, Li$$_{2}$$ZrO$$_{3}$$, is known as a candidate blanket material in a fusion reactor. According to the independent benchmark studies for zirconium by JAERI, Kyoto University and Osaka University, the neutron spectrum calculations show fairly large overestimation for most evaluated nuclear data libraries. The author's group expects that the overestimation be due to a problem of evaluation for the (n,2n) reaction, because the (n,2n) reaction cross section is not well determined experimentally. In the present study, two neutrons emitted from $$^{rm nat}$$Zr(n,2n) reaction have been measured directly to reveal the problem. As a result of measurements, the cross section obtained for energies above 1 MeV, which is the lower measurable limit energy, shows a little larger than JENDL-3.3. This is an opposite result to the benchmark analysis. However, an extrapolation for the low energy region by the evaporation spectrum with the nuclear temperature of 1 MeV brought the smaller total (n,2n) reaction cross section than JENDL-3.3, which is comparable to ENDF/B-VI. This result suggests that the discrepancies reported previously might be due to inappropriate evaluation of nuclear temperature.

Journal Articles

Torus configuration and materials selection on a fusion DEMO reactor, SlimCS

Tobita, Kenji; Nishio, Satoshi; Tanigawa, Hiroyasu; Enoeda, Mikio; Isono, Takaaki; Nakamura, Hirofumi; Tsuru, Daigo; Suzuki, Satoshi; Hayashi, Takao; Tsuchiya, Kunihiko; et al.

Journal of Nuclear Materials, 386-388, p.888 - 892, 2009/04

 Times Cited Count:25 Percentile:83.46(Materials Science, Multidisciplinary)

SlimCS is the conceptual design of a compact fusion DEMO plant assuming technologies foreseeable in 2020s-2030s. Considering continuity of blanket technology from the Japanese proposal on ITER-TBM, the prime option of blanket is water-cooled solid breeder with Li$$_{2}$$TiO$$_{3}$$ and Be (or Be$$_{12}$$Ti). A reduced-activation ferritic-martensitic steel and pressurized water are chosen as the structural material and coolant, respectively. Toroidal coils produce the peak magnetic field above 16 T using the RHQT processed Nb$$_{3}$$Al conductors. The structure and materials of the conducting shell and divertor are also presented.

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