論文 - 星野　一生

Photon Trapping Effects in DEMO Divertor Plasma

Hoshino, K., Sawada, K., Idei, R., Tokunaga, S., Asakura, N., Shimizu, K., Ohno, N.

CONTRIBUTIONS TO PLASMA PHYSICS 56 （ 6-8 ） 657 - 662 2016年08月

研究論文（国際会議プロシーディングス）, 共著, 筆頭著者, 査読有り,  ISSN  0863-1042

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Effects of Classical and Neo-classical Cross-field Transport of Tungsten Impurity in Realistic Tokamak Geometry

Yamoto, S., Homma, Y., Inoue, H., Sawada, Y., Hoshino, K., Hatayama, A., Bonnin, X., Coster, D., Schneider, R.

CONTRIBUTIONS TO PLASMA PHYSICS 56 （ 6-8 ） 646 - 650 2016年08月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0863-1042

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Particle pinch model of passing/trapped high-Z impurity with centrifugal force effect

Shimizu Yusuke, Fujita Takaaki, Okamoto Atsushi, Arimoto Hideki, Hayashi Nobuhiko, Hoshino Kazuo, Nakano Tomohide, Honda Mitsuru

Plasma and Fusion Research 11   2403082 2016年06月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  1880-6821

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<p>The inward pinch model for high-Z impurity ions in a toroidally rotating plasma has been improved to take account of trapped particles and the centrifugal force effect. It is found that the PHZ pinch is large for the barely trapped particles and the pinch velocity increases with the rotation velocity as in the original model.</p>

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Self-consistent treatment of the sheath boundary conditions by introducing anisotropic ion temperatures and virtual divertor model

Togo Satoshi, Takizuka Tomonori, Nakamura Makoto, Hoshino Kazuo, Ibano Kenzo, Lang Tee Long, Ogawa Yuichi

Journal of Computational Physics 310   109 - 126 2016年04月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  0021-9991

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<p>One-dimensional SOL-divertor plasma fluid simulation code which considers anisotropy of ion temperature has been developed so as to deal with sheath theory self-consistently. In our fluid modeling, explicit use of boundary condition for Mach number M at divertor plate, e.g., M=1, becomes unnecessary. In order to deal with the Bohm condition and the sheath heat transmission factors at divertor plate self-consistently, we introduced a virtual divertor (VD) model which sets an artificial region beyond divertor plates and artificial sinks for particle, momentum and energy there to model the effects of the sheath region in front of the divertor plate. Validity of our fluid model with VD model is confirmed by showing that simulation results agree well with those from a kinetic code regarding the Bohm condition, ion temperature anisotropy and supersonic flow. We also show that the strength of artificial sinks in VD region does not affect profiles in plasma region at least in the steady state and that sheath heat transmission factors can be adjusted to theoretical values by VD model. Validity of viscous flux is also investigated.</p>

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Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

Utoh Hiroyasu, Takase Haruhiko, Sakamoto Yoshiteru, Tobita Kenji, Mori Kazuo, Kudo Tatsuya, Someya Youji, Asakura Nobuyuki, Hoshino Kazuo, Nakamura Makoto, Tokunaga Shinsuke

Fusion Engineering and Design 103   93 - 97 2016年02月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0920-3796

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<p>In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell.</p>

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Design study of blanket structure based on a water-cooled solid breeder for DEMO

Someya Youji, Tobita Kenji, Utoh Hiroyasu, Tokunaga Shinji, Hoshino Kazuo, Asakura Nobuyuki, Nakamura Makoto, Sakamoto Yoshiteru

Fusion Engineering and Design 98-99   1872 - 1875 2015年10月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0920-3796

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<p>Blanket concept with a simplified interior for mass production has been developed using a mixed bed of Li₂TiO₃ and Be₁₂Ti pebbles, coolant conditions of 15.5 MPa and 290-325 °C and cooling pipes without any partitions. Considering the continuity with the ITER test blanket module option of Japan and the engineering feasibility in its fabrication, our design study focused on a water-cooled solid breeding blanket using the mixed pebbles bed. Herein, we propose blanket segmentation corresponding to the shape and dimension of the blanket and routing of the coolant flow. Moreover, we estimate the overall tritium breeding ratio (TBR) with a torus configuration, based on the segmentation using three-dimensional (3D) Monte Carlo N-particle calculations. As a result, the overall TBR is 1.15. Our 3D neutronics analysis for TBR ensures that the blanket concept can achieve a self-sufficient supply of tritium.</p>

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Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

Utoh Hiroyasu, Tobita Kenji, Someya Youji, Asakura Nobuyuki, Sakamoto Yoshiteru, Hoshino Kazuo, Nakamura Makoto

Fusion Engineering and Design 98-99   1648 - 1651 2015年10月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0920-3796

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<p>Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) 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. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.</p>

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Effects of background plasma characteristics on tungsten impurity transport in the SOL/divertor region using IMPGYRO code

Yamoto, S., Homma, Y., Hoshino, K., Sawada, Y., Bonnin, X., Coster, D., Schneider, R., Hatayama, A.

JOURNAL OF NUCLEAR MATERIALS 463   615 - 619 2015年08月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0022-3115

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Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

Takizuka, T., Tokunaga, S., Hoshino, K., Shimizu, K., Asakura, N.

JOURNAL OF NUCLEAR MATERIALS 463   1229 - 1232 2015年08月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0022-3115

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Simulation study of power load with impurity seeding in advanced divertor "short super-X divertor" for a tokamak reactor

Asakura, N., Hoshino, K., Shimizu, K., Shinya, K., Utoh, H., Tokunaga, S., Tobita, K., Ohno, N.

JOURNAL OF NUCLEAR MATERIALS 463   1238 - 1242 2015年08月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0022-3115

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The influence of the radial particle transport on the divertor plasma detachment

Hoshino, K., Shimizu, K., Takizuka, T., Asakura, N., Nakano, T.

JOURNAL OF NUCLEAR MATERIALS 463   573 - 576 2015年08月

研究論文（国際会議プロシーディングス）, 共著, 筆頭著者, 査読有り,  ISSN  0022-3115

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SOL-divertor plasma simulations introducing anisotropic temperature with virtual divertor model

Togo, S., Takizuka, T., Nakamura, M., Hoshino, K., Ogawa, Y.

JOURNAL OF NUCLEAR MATERIALS 463   502 - 505 2015年08月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0022-3115

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Analysis of Tungsten Transport in JT-60U Plasmas

Y.Shimizu, T.Fujita, H.Arimoto, T.Nakano, K.Hoshino, N.Hayashi

Plasma Fusion Research 10   3403062 2015年07月

研究論文（学術雑誌）, 共著, 査読有り

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Relationship between net electric power and radial build of DEMO based on ITER steady-state scenario parameters

Sakamoto, Y., Nakamura, M., Tobita, K., Utoh, H., Someya, Y., Hoshino, K., Asakura, N., Tokunaga, S.

FUSION ENGINEERING AND DESIGN 89 （ 9-10 ） 2440 - 2445 2014年10月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0920-3796

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Application of inter-linked superconducting coils for central solenoid and advanced divertor configuration of DEMO

Utoh Hiroyasu, Asakura Nobuyuki, Tobita Kenji, Sakamoto Yoshiteru, Someya Youji, Hoshino Kazuo, Nakamura Makoto

Fusion Engineering and Design 89 （ 9-10 ） 2456 - 2460 2014年10月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0920-3796

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<p>To determine the engineering feasibility of an "inter-linked" (IL) superconducting central solenoid (CS) and poloidal field (PF) coil, we studied the basic engineering design of the IL-CS and PF coil. We were able to use IL superconducting coils in a tokamak fusion demonstration power plant (DEMO) reactor to achieve a sufficient amount of the CS magnetic flux swing for the current ramp-up while maintaining plasma size and also achieving an advanced divertor concepts of "super-X divertor (SXD)" by locating near the plasma. A basic idea of the IL superconducting CS and PF coil concept is to wind a CS (PF coil) such that it is linked in a set of toroidal field (TF) coils. In this study, we considered the conductor and feeder design of the IL-CS and PF coil including the superconducting strand type, cooling, and winding and assembly methods, and determined that react and wind Nb₃Al is a candidate for use in an IL coil and feeder in this specification. This paper presents the preliminary engineering design results of the IL superconducting CS and PF coil for creating a SXD equilibrium configuration.</p>

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Waste management scenario in the hot cell and waste storage for DEMO

Someya Youji, Tobita Kenji, Yanagihara Satoshi, Kondo Masatoshi, Utoh Hiroyasu, Asakura Nobuyuki, Hoshino Kazuo, Nakamura Makoto, Sakamoto Yoshiteru

Fusion Engineering and Design 89 （ 9-10 ） 2033 - 2037 2014年10月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0920-3796

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<p>Management strategy of radioactive waste generated in periodic replacement may be important in the point of view of fusion reactor design, because it has a large impact on the design of the hot cell and waste storage in the plant. In the replacement period of a fusion power reactor, the assembly of blanket or divertor modules needs to be removed from the reactor in order to minimize remote maintenance in the vacuum vessel and to attain reasonable plant availability. In the hot cell, the modules will be removed from the back plate 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. On the other hand, it is assumed that the structural material (F82H) 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. In this paper, the basic idea of the waste management scenario and the conceptual design in the hot cell and waste storage for DEMO has been proposed.</p>

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Development of the Backflow Model for Simplified Impurity Exhaust in Monte-Carlo Calculation

K. Hoshino, K. Shimizu, H. Kawashima, T. Takizuka, T. Nakano, S. Ide

Contributions to Plasma Physics 54   404 - 408 2014年06月

研究論文（国際会議プロシーディングス）, 共著, 筆頭著者, 査読有り

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Divertor study on DEMO reactor

Hoshino Kazuo, Asakura Nobuyuki, Shimizu Katsuhiro, Tokunaga Shinsuke, Takizuka Tomonori, Someya Youji, Nakamura Makoto, Utoh Hiroyasu, Sakamoto Yoshiteru, Tobita Kenji

Plasma and Fusion Research 9   3403070 2014年06月

研究論文（学術雑誌）, 共著, 筆頭著者, 査読有り,  ISSN  1880-6821

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<p>Huge power handling in the SOL/divertor region is one of the crucial issues for a tokamak fusion reactor. Divertor design study of a DEMO reactor with fusion power of 3 GWand ITER size plasma has progressed using the integrated divertor code SONIC. Recently, to improve conversion of the solution for the DEMO divertor plasma simulation, SONIC code has been improved. The calculation time is significantly reduced by (i) the backflow model for the simplified impurity exhaust process and (ii) optimization on HELIOS at BA-IFERC. In the SONIC simulation, the partial detached divertor plasma was obtained by the Ar impurity seeding. Although the plasma heat load at the outer target was reduced by the partial detachment, the contribution of the impurity radiation and the surface recombination of the fuel ions to the target heat load became large. As a result, the peak of the total target heat load was estimated to be 16MW/m². In order to reduce the total heat load, control of the impurity radiation profile by kind of seeding impurity species and the divertor geometry has been studied. They can decrease the target heat load, but the peak heat load is still larger than the heat removal capability of the present divertor target concept. Further design study including change of the machine specifications is necessary.</p>

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Systematic Study of Tungsten Impurity Transport in Representative Regimes of Divertor Plasma

Yamoto, S., Hoshino, K., Toma, M., Homma, Y., Hatayama, A., Bonnin, X., Coster, D., Schneider, R.

CONTRIBUTIONS TO PLASMA PHYSICS 54 （ 4-6 ） 421 - 425 2014年06月

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  0863-1042

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A simulation study of large power handling in the divertor for a Demo reactor

Asakura Nobuyuki, Shimizu Katsuhiro, Hoshino Kazuo, Tobita Kenji, Tokunaga Shinsuke, Takizuka Tomonori

Nuclear Fusion 53 （ 12 ） 123013 2013年12月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  0029-5515

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<p>Power exhaust for a 3 GW class fusion reactor with an ITER-sized plasma was investigated by enhancing the radiation loss from seeding impurity. The impurity transport and plasma detachment were simulated under the Demo divertor condition using an integrated divertor code SONIC, in which the impurity Monte-Carlo code, IMPMC, can handle most kinetic effects on the impurity ions in the original formula. The simulation results of impurity species from low Z (neon) to high Z (krypton) and divertor length with a plasma exhausted power of 500 MW and radiation loss of 460 MW, and a fixed core-edge boundary of 7 × 10¹⁹ m^-3 were investigated at the first stage for the Demo divertor operation scenario and the geometry design. Results for the different seeding impurities showed that the total heat load, including the plasma transport and radiation , was reduced from 15-16 MW m^-2 (Ne and Ar) to 11 MW m^-2 for the higher Z (Kr), and extended over a wide area accompanied by increasing impurity recycling. The geometry effect of the long-leg divertor showed that full detachment was obtained, and the peak qtarget value was decreased to 12 MW m^-2, where neutral heat load became comparable to and due to smaller flux expansion. Fuel dilution was reduced but was still at a high level. These results showed that a divertor design with a long leg with higher Z seeding such as Ar and Kr is not fulfilled, but will be appropriate to obtain the divertor scenario for the Demo divertor. Finally, influences of χ and D enhancement were seen significantly in the divertor, i.e. the radiation and density profiles became wider, leading to full detachment. Both qtarget near the separatrix and Te at the outer flux surfaces were decreased to a level for the conventional technology design. On the other hand, the problem of fuel dilution became worse. Extrapolation of the plasma transport coefficients to ITER and Demo, where density and temperature will be higher than ITER and edge-localized modes are mitigated, is a key issue for the divertor design.</p>

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