Ishizawa, Akihiro*; Idomura, Yasuhiro; Imadera, Kenji*; Kasuya, Naohiro*; Kanno, Ryutaro*; Satake, Shinsuke*; Tatsuno, Tomoya*; Nakata, Motoki*; Nunami, Masanori*; Maeyama, Shinya*; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 92(3), p.157 - 210, 2016/03
The high-performance computer system Helios which is located at The Computational Simulation Centre (CSC) in The International Fusion Energy Research Centre (IFERC) started its operation in January 2012 under the Broader Approach (BA) agreement between Japan and the EU. The Helios system has been used for magnetised fusion related simulation studies in the EU and Japan and has kept high average usage rate. As a result, the Helios system has contributed to many research products in a wide range of research areas from core plasma physics to reactor material and reactor engineering. This project review gives a short catalogue of domestic simulation research projects. First, we outline the IFERC-CSC project. After that, shown are objectives of the research projects, numerical schemes used in simulation codes, obtained results and necessary computations in future.
Uto, Hiroyasu; Takase, Haruhiko; Sakamoto, Yoshiteru; Tobita, Kenji; Mori, Kazuo; Kudo, Tatsuya; Someya, Yoji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; et al.
Fusion Engineering and Design, 103, p.93 - 97, 2016/02
Conceptual design of in-vessel component including conducting shell has been investigated in Broader Approach (BA) DEMO design activities, in order to propose feasible DEMO reactor from plasma vertical stability and engineering viewpoint. The conducting shell for the plasma vertical stability will be incorporated behind blanket module, while the location must be close to the plasma surface as possible for the plasma stabilization. We evaluated dependence of the plasma vertical stability on the conducing shell parameters by using a 3-dimensional eddy current analysis code (EDDYCAL). The calculation results showed that the conducting shell requires more than 0.01 m thickness of Cu-alloy on DEMO. On the other hand, the electromagnetic force at the plasma disruption is a few times larger than no conducting shell case because of larger eddy current on conducting shell. The engineering design issues of in-vessel components for plasma vertical stability are presented.
Hoshino, Kazuo; Matsunaga, Go; Okumura, Yoshikazu
Purazuma, Kaku Yugo Gakkai-Shi, 92(2), p.146 - 147, 2016/02
no abstracts in English
Hoshino, Kazuo; Matsunaga, Go; Okumura, Yoshikazu
Purazuma, Kaku Yugo Gakkai-Shi, 91(12), p.802 - 803, 2015/12
Holding of the 17th IFERC Project Committee (PC), the 16th IFMIF/EVEDA-PC and the progress of the JT-60SA were reported. In the IFERC-PC, total 33 participated including Maisonnier chairperson, JA-EU committee, experts and project members. They reported the status of each activity, discussed the planning of project 2016 and revised plan, and concluded technical advice to the BA Steering Committee (SC). In the IFMIF/EVEDA-PC, total 27 participated including Takatsu chairperson, JA-EU committee, experts and project members. They reported the status of each activity, discussed the planning of project 2016 and revised plan, and the BA-SC recommended the endorsement of the new project plan. In the JT-60SA, a rotary crane for assembly of the troidal field coil has been installed. The disassembly of the constraint jigs of the vacuum vessel which completed 340 assembly has been progressed. 6 high temperature superconductor current leads for the troidal field coils had been delivered from EU.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Tokunaga, Shinsuke; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru
Fusion Engineering and Design, 98-99, p.1872 - 1875, 2015/10
Blanket concept with simplified interior for mass production has been developed with a mixed bed of LiTiO and BeTi pebbles, a coolant condition of 15.5 MPa and 290-325C and cooling tubes only without any partitions. A neutronics analysis ensured the blanket concept meets a self-sufficient supply of tritium. However, this concept is vulnerable to the inner pressure. A plant availability for DEMO may drop to a lower value, because a potential of resume operations after an accident such as a coolant leakage in blanket is not considered. The blanket design will be revisited for the availability. Considering the continuity with the ITER-TBM option of Japan and the engineering feasibility of fabrication, our design study focuses on a water-cooled solid breeding blanket using the mixed pebbles bed. A breakage of the blanket casing should be avoided not to contaminate the plasma chamber with water and breeding materials. A water-cooled solid blanket with inner pressure tightness is estimated by the ANSYS code. As a results, the pressure tightness of 8 MPa (water vapor pressure at 300C) can be compatible with the self-sufficient production of tritium when the blanket is as thick as about 0.9 m and the ribs are arranged in the radial direction. Therefore, the blanket concept with pressure tightness of 8 MPa is adopted with depressurization system as which a tritium recovery system such as helium purge-gas line is posteriorly arranged in blanket to serve. On the other hand, a handling of decay heat is a serious problem at an accident such as LOCA. Coolant flow is divided into the blanket to secure heat removal for the safety. Finally, the blanket segmentation with the shape and dimension of blanket and routing of coolant flow has also been proposed. Moreover, overall TBR is estimated with torus configuration based in the segmentation using three-dimensional MCNP calculation.
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto
Fusion Engineering and Design, 98-99, p.1648 - 1651, 2015/10
Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field coil, the arrangement of poloidal field coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. In this study, we categorize various schemes in term of (1) the maintenance port position for transporting blanket segments, (2) blanket segmentation, and (3) divertor segmentation. In reviewing these assessment factors, the separated sector transport using the vertical maintenance ports with small divertor cassette maintenance scheme was found to be a more probable maintenance approach. This presentation describes engineering design of each maintenance schemes and evaluation results of comparison among maintenance schemes.
Hoshino, Kazuo; Matsunaga, Go; Okumura, Yoshikazu
Purazuma, Kaku Yugo Gakkai-Shi, 91(10), p.700 - 701, 2015/10
The progress of IFMIF Prototype Accelerator in the IFERC site and the Satellite Tokamak (JT-60SA) project were reported. For the IFMIF Prototype Accelerator, two radio frequency (RF) sources of the RF quadrupole accelerator, seven high-voltage power supplies, and distribution cabinets were delivered from the CIEMAT institute in Spain and the installation was started. In the Satellite Tokamak project, the development for long term operation technique of the neutral beam injector has been achieved and thereby confirmed to be feasible for long term operation of JT-60SA. The 340 torus assembly of vacuum vessel of the JT-60SA, which was started from May last year, has been completed. To send out information to the domestic community a major event in the BA activities.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke
Fusion Science and Technology, 68(2), p.423 - 427, 2015/09
The radioactive waste is generated in every replacement of an in-vessel component. Maintenance scheme is to replace the blanket segment and divertor cassette independently, as the lifetime of them is different. The blanket segment consists of some blanket modules mounted to back-plate. Total weight is estimated to amount to about 6,648 ton (1,575 ton of blanket module, 3,777 ton of back-plate, 372 ton of conducting shell and 924 ton of divertor cassette). In base case, main parameters of DEMO reactor are 8.2 m of major radius and 1.35 GW of fusion output. The lifetimes of blanket segment and divertor cassette are assumed to be 2.2 years and 0.6 year, respectively, 52,487 ton wastes is generated in plant life of 20 years. Therefore, there is a concern that a contamination controlled area for the radioactive waste may increase because much the waste is generated in every replacement. In this paper, management scenario is proposed to reduce the radioactive waste. The back-plates and cassette bodies (628 ton) of divertor was reused. As a result, the displacement per atom (DPA) of the back-plates of SUS316L was 0.2 DPA/year and that of the cassette bodies of F82H was 0.6 DPA/year. Therefore, reusing the back-plates and cassette bodies would be possible, if re-welding points are arranged under neutron shielding. It was found that radioactive waste could be reduced to 20%, when tritium breeding materials are recycled. Finally, a design of DEMO building such as a hot cell and temporary storage etc. is proposed.
Asakura, Nobuyuki; Hoshino, Kazuo; Shimizu, Katsuhiro; Shinya, Kichiro*; Uto, Hiroyasu; Tokunaga, Shinsuke; Tobita, Kenji; Ono, Noriyasu*
Journal of Nuclear Materials, 463, p.1238 - 1242, 2015/08
Arrangements of interlink divertor coils and divertor geometries for short super-X was proposed as the Demo advanced divertor design. Performance of plasma detachment under the large heat flux was investigated to optimize the divertor design, using SONIC simulation with Ar impurity seeding, where Pout = 500 MW, ne = 710 m at the core-edge boundary and the same diffusion coefficients for ITER simulation. Effects on the plasma temperature and density distributions were compared to the conventional divertor. The first run results with the same radiation power fraction of 0.92 in the conventional divertor showed that full detached plasma is produced, the maximum radiation region was maintained upstream the divertor target, and both the plasma heat load plus radiation load at the target was reduced to 10 MWmm level. Simulation for the lower radiation power fractions of 0.8-0.9 was also performed, and physics issues of the short super-X divertor are discussed.
Hoshino, Kazuo; Shimizu, Katsuhiro; Takizuka, Tomonori*; Asakura, Nobuyuki; Nakano, Tomohide
Journal of Nuclear Materials, 463, p.573 - 576, 2015/08
Togo, Satoshi*; Takizuka, Tomonori*; Nakamura, Makoto; Hoshino, Kazuo; Ogawa, Yuichi*
Journal of Nuclear Materials, 463, p.502 - 505, 2015/08
A 1D SOL-divertor plasma simulation code introducing the anisotropic ion temperature with virtual divertor model has been developed. By introducing the anisotropic ion temperature directly, the second-derivative parallel ion viscosity term in the momentum transport equation can be excluded and the boundary condition at the divertor plate becomes unnecessary. In order to express the effects of the divertor plate and accompanying sheath implicitly, the virtual divertor model has been introduced which has an artificial sinks of particle, momentum and energy. The virtual divertor model makes the periodic boundary condition available. By using this model, SOL-divertor plasmas satisfying the Bohm condition has been successfully obtained. Also investigated are the dependence of the ion temperature anisotropy on the normalized mean free path of ion and the validity of the approximated parallel ion viscosity for the Braginskii expression and the limited one.
Hoshino, Kazuo; Matsunaga, Go; Okumura, Yoshikazu
Purazuma, Kaku Yugo Gakkai-Shi, 91(8), p.561 - 562, 2015/08
The new construction of Joint Research Building for the DEMO R&D at IFERC, progress of the IFMIF Prototype Accelerator and the Satellite Tokamak (JT-60SA) project were reported. The Joint Research Building processing the construction as a new place where carry out a part of project for the DEMO R&D in the rest of the BA activity period and it mainly consists of the laboratory/material testing room for the material evaluation and Be production room for handling of the metal beryllium. For the IFMIF Prototype Accelerator, completed the repairs and test run of the Interlock, successful of the continuity rated proton beam, then after setting of the radiation control area, successful production of the deuterium ion beam. For the JT-60SA, the installation of a Helium refrigerator which is one of the world's largest and a He buffer tank have been completed. The 4th Research Coordination meeting was held and they discussed the research plan and theme for the experiment proposal of JT-60SA.
Hayashi, Nobuhiko; Honda, Mitsuru; Shiraishi, Junya; Miyata, Yoshiaki; Wakatsuki, Takuma; Hoshino, Kazuo; Toma, Mitsunori; Suzuki, Takahiro; Urano, Hajime; Shimizu, Katsuhiro; et al.
Europhysics Conference Abstracts (Internet), 39E, p.P5.145_1 - P5.145_4, 2015/06
Hoshino, Kazuo; Matsunaga, Go; Okumura, Yoshikazu
Purazuma, Kaku Yugo Gakkai-Shi, 91(6), p.430 - 431, 2015/06
Holding of the 16th BA Steering Committee and progress of the Satellite Tokamak (JT-60SA) project were reported. There were 33 participants for the 16th BA Steering Committee, including 3 committee members and 10 specialists from EU, 4 committee and 11 specialists from JP, each project leaders and project committee chair persons, and they approved for update of the 2014 Annual Report and the Work Plan of IFMIF/EVEDA, the IFERC and the Satellite Tokamak project. In the construction of JT-60SA, the component carry-in from EU and the assembly has been progressed steadily in JP. To announce for these progressing, a ceremony was held and 200 people concerned in EU-JP attended. The 22nd JT-60SA Technical Coordination Meeting was held. 73 people EU-JP joined and had discussion regarding the design, status of production, examination of assembly and the integrate operation in the future. To send out information to the domestic community a major event in the BA activities.
Someya, Yoji; Tobita, Kenji; Tanigawa, Hisashi; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
This paper presents neutronics analysis mainly focused on key design issues for self-sufficient tritium production based on the conceptual design study carried out for a fusion DEMO reactor in past several years, which includes new findings regarding design methodology of breeding blanket. Self-sufficient production of tritium is one of the most critical requirements for fusion reactors. We considered a fusion DEMO reactor with a major radius of about 8 m and fusion output of 1.5 GW with breeding blanket consisting of a mixed bed of LiTiO and BeTi pebbles. The net tritium breeding ratio (TBR) was estimated to be 1.15 with a three-dimensional analysis with the MCNP-5 with nuclear library of FENDL-2.1, satisfying a self-sufficient supply of tritium (net TBR1.05). Throughout the research, we found that tritium breeding capability (i.e., local TBR) of breeding blanket is less dependent on the arrangement of cooling pipe in the blanket when the neutron wall loading is lower than about 1.5 MW/m which is met in the DEMO considered. The result suggests that tolerance for the installation of cooling pipes in each blanket module will not be a critical matter. In addition, we found that a gap of about 0.02 m between neighboring blanket modules has little impact on the gross TBR.
Sakamoto, Yoshiteru; Nakamura, Makoto; Tobita, Kenji; Uto, Hiroyasu; Someya, Yoji; Hoshino, Kazuo; Asakura, Nobuyuki; Tokunaga, Shinsuke
Fusion Engineering and Design, 89(9-10), p.2440 - 2445, 2014/10
Several concepts of DEMO have been proposed so far with plasma physics assumptions. At the same time, plasma performances foreseen in DEMO have been developed experimentally in tokamaks. However there are large gaps between the physics design parameters of the DEMO concepts and the simultaneous achieved parameters in tokamak experiments. Since one of the foreseeable integrated plasma performances is the ITER steady-state scenario, the projection of the scenario parameter to DEMO concept has been analyzed by using the systems code. The fusion power of 1GW can be obtained with the plasma major radius of 9 m. The same power can be obtained with 8 m if the distance between TF coil and plasma surface is reduced from 2 m to 1.5 m. Furthermore, it was found that the heat load on the divertor region is increased with increasing the normalized density and is decreased with increasing the normalized beta.
Someya, Yoji; Tobita, Kenji; Yanagihara, Satoshi*; Kondo, Masatoshi*; Uto, Hiroyasu; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Sakamoto, Yoshiteru
Fusion Engineering and Design, 89(9-10), p.2033 - 2037, 2014/10
In the replacement period of a fusion power reactor, the assembly of blanket or divertor modules need to be removed from the reactor in order to minimize remote maintenance in the vacuum vessel and to attain a reasonable plant availability. In the hot cell, the modules will be removed from the backplate of the assembly. Here, note that the active cooling must be done by a way that does not cause contamination of the hot cell environment due to dispersion of tritium and tungsten dust. In this sense, the cooling scenario is adopted that the existing pipe of cooling water in the assembly is connected to a different cooling water system in the hot cell. In this scenario, the temperature of the assembly is maintained about 40-100C. On the other hand, the structural material (RAFM) of the blanket and divertor is not recycled due to its high contact dose rate. It should be crushed into small pieces to reduce volume of the waste and required storage space. Here, the decay heat must be removed by natural convection to keep the temperature below 65C for preventing water evaporation from the mortar. The RAFM is kept in the interim storage during 12 years until the required temperature conditions for mortar are ensured and then is disposed of.
Uto, Hiroyasu; Asakura, Nobuyuki; Tobita, Kenji; Sakamoto, Yoshiteru; Someya, Yoji; Hoshino, Kazuo; Nakamura, Makoto
Fusion Engineering and Design, 89(9-10), p.2456 - 2460, 2014/10
Recently, use of an inter-linked (IL) superconducting coils in a tokamak fusion DEMO reactor were proposed. A basic idea of the IL-CS concept is to wind a CS such that it is linked in a set of toroidal field (TF) coils. In this presentation, the detailed descriptions of the engineering design of the superconducting CS linked in TFCs will be presented. Handling of a large exhausted power from the core plasma is the most important issue for the fusion reactor. Recently, advanced divertor concepts of super-X divertor (SXD) was proposed. The plasma equilibrium calculations for SlimCS showed that large coil currents are required for the conventional divertor coil location outside TFC. These results show that installation of the divertor coils inter-TFC (inter-linked PF) is required for the DEMO advanced divertor design. In this presentation, engineering feasibility of the inter-linked superconducting CS and PF for constructing the SXD equilibrium configuration will be presented.
Hoshino, Kazuo; Asakura, Nobuyuki; Shimizu, Katsuhiro; Tokunaga, Shinsuke
Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 6 Pages, 2014/10