論文 - 星野　一生

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

　概要を見る

<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

　概要を見る

<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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Critical design factors for sector transport maintenance in DEMO

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

Nuclear Fusion 53 （ 12 ） 123005 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>This paper mainly focuses on a sector transport maintenance scheme from the aspects of high plant availability. In this study, three different maintenance schemes are considered based on (1) the number of maintenance ports and (2) the insertion direction. The design study clarifies critical design factors and key engineering issues on the maintenance scheme: (1) how to support an enormous overturning force of the toroidal field coils in the large open port for sector transport and (2) define the transferring mechanism of sectors in the vacuum vessel. On reviewing these assessment factors, the sector transport using a limited number of horizontal maintenance ports is found to be a more realistic maintenance scheme. In addition, evaluating maintenance scenarios under high decay heat is proposed for the first time. The key design factors are the cool-down time in the reactor and the cooling method in the maintenance scheme to keep components under operational temperature. Based on one-dimensional heat conduction analysis, after one month cool-down time, each sector of SlimCS could be transported to the hot cell facility by gas cooling.</p>

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Effects of Neutral Particles on the Stability of the Detachment Fronts in Divertor Plasmas

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

Plasma Fusion Research 8   2403096 2013年08月

研究論文（学術雑誌）, 共著, 査読有り

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Comparison of kinetic and fluid models for tungsten impurity transport using IMPGYRO and SOLPS

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

JOURNAL OF NUCLEAR MATERIALS 438   S620 - S624 2013年07月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  0022-3115

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Investigation of advanced divertor magnetic configuration for DEMO tokamak reactor

N. Asakura, K. Shinya, K. Tobita, K. Hoshino, K. Shimizu, H. Utoh, Y. Someya, M. Nakamura, N. Ohno, M. Kobayashi, H. Tanaka

Fusion Science and Technology 63   70 - 75 2013年03月

研究論文（国際会議プロシーディングス）, 共著, 査読有り

Effects of radial losses of particle and energy on the stability of detachment front in a divertor plasma

Togo Satoshi, Nakamura Makoto, Ogawa Yuichi, Shimizu Katsuhiro, Takizuka Tomonori, Hoshino Kazuo

Plasma and Fusion Research 7   2401087 2012年12月

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

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<p>Operation under partially detached divertor (PDD) plasmas is a hopeful way in order to reduce the divertor heat load in the next generation tokamaks. The physical mechanism of PDD plasmas, however, has not fully been understood yet. We have studied them with a multi-layer one-dimensional divertor model. The PDD plasmas are successfully reproduced by introducing a neutral gas puffing model. Effect of the cross-field heat transport on the PDD plasmas is investigated. It is found that cross-field heat transport both in the SOL region and in the divertor region prevents detachment fronts from moving upstream in a detached flux tube.</p>

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Simulation Study of L/H Transition with Self-Consistent Integrated Modelling of Core and SOL/Divertor Transport

Yagi, M., Shimizu, K., Takizuka, T., Honda, M., Hayashi, N., Hoshino, K., Fukuyama, A.

CONTRIBUTIONS TO PLASMA PHYSICS 52 （ 5-6 ） 372 - 378 2012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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Simulation Study of an Extended Divertor Leg for Heat Control in the SlimCS DEMO Reactor

Hoshino, K., Shimizu, K., Asakura, N., Takizuka, T., Nakamura, M., Tobita, K.

CONTRIBUTIONS TO PLASMA PHYSICS 52 （ 5-6 ） 550 - 554 2012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,  ISSN  0863-1042

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Development of Coupled IMPGYRO-SOLPS Codes for Analyzing Tokamak Plasmas with Tungsten Impurities

Toma, M., Bonnin, X., Hoshino, K., Hatayama, A., Schneider, R., Coster, D.

CONTRIBUTIONS TO PLASMA PHYSICS 52 （ 5-6 ） 450 - 454 2012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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