論文 - 星野　一生

Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

Asakura N., Hoshino K., Suzuki S., Tokunaga S., Someya Y., Utoh H., Kudo H., Sakamoto Y., Hiwatari R., Tobita K., Shimizu K., Ezato K., Seki Y., Ohno N., Ueda Y.

Nuclear Fusion 57 （ 12 ） 126050 2017年10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 to the divertor and the conceptual design have been investigated for a steadystate DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of P_sep = 205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of P_out= 250 MW and the total radiation fraction at the edge, SOL and divertor (P_rad/ P_out = 0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load (q_target) at the attached region was reduced to∼5 MW m^-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak q_target was less than 10 MW m^-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak q_target of 10 MWm^-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m^-2 was distributed in the major part of the coolant pipe. These results were acceptable for the plasma facing and structural materials.</p>

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Numerical analysis of tungsten erosion and deposition processes under a DEMO divertor plasma

Homma Yuki, Hoshino Kazuo, Yamoto Shohei, Asakura Nobuyuki, Tokunaga Shinsuke, Hatayama Akiyoshi, Sakamoto Yoshiteru, Hiwatari Ryoji, Tobita Kenji, DEMO Joint Special Design Team for Fusion

Nuclear Materials and Energy 12   323 - 328 2017年09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  2352-1791

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<p>Erosion reduction of tungsten (W) divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density n _i and higher temperature case and the other is higher n _i and lower temperature case. Net erosion has been seen in the lower n _i case. But in the higher n _i case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors are important to understand the net erosion formation: (i) ratio of the 1st ionization length of sputtered W atom to the Larmor gyro radius of W+ ion, (ii) balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.</p>

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Development of bolometric tomography technique for high-contrast radiation distribution in toroidal devices

Sano Ryuichi, Mukai Kiyofumi, Peterson Byron J., Fukumoto Masakatsu, Hoshino Kazuo

Review of Scientific Instruments 88 （ 9 ） 093504 2017年09月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  0034-6748

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<p>An extended Computed Tomography (CT) technique with a priori information based on Phillips-Tikhonov regularization has been developed to handle a high-contrast radiation distribution, which can result in large reconstruction errors in the region where radiation intensity is low, with few line of sight (LOS) data. Reference profiles generated from LOS data for every time slice are employed as the a priori information. In the extended technique, the weighting parameter for the reference profile is automatically determined from the LOS data, to avoid an inappropriate reference. The extended technique has been examined with a reconstruction test with the previously designed CT system using an infrared imaging video bolometer in JT-60SA. In the reconstructed profile of the high-contrast radiation distribution, the extended technique shows improvement of the weak radiation region, such as the scrape-off layer, without the deterioration of the strong radiation region, such as around the divertor. The results indicate that the extended CT technique can help to handle a high-contrast radiation distribution with few LOS data.</p>

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Kinetic modeling of high-Z tungsten impurity transport in ITER plasmas using the IMPGYRO code in the trace impurity limit

Yamoto S., Bonnin X., Homma Y., Inoue H., Hoshino K., Hatayama A., Pitts R. A.

Nuclear Fusion 57 （ 11 ） 116051 2017年08月

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

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<p>In order to obtain a better understanding of tungsten (W) transport processes, we are developing the Monte-Carlo W transport code IMPGYRO. The code has the following characteristics which are important for calculating W transport: (1) the exact Larmor motion of W ions is computed so that the effects of drifts are automatically taken into account; (2) Coulomb collisions between W impurities and background plasma ions are modelled using the Binary Collision Model which provides more precise kinetic calculations of the friction and thermal forces. By using the IMPGYRO code, the W production/transport in the ITER geometry has been calculated under two different divertor operation modes (Case A: partially detached state and Case B: high recycling state) obtained from the SOLPS-ITER code suite calculation without the effect of drifts. The results of the W-density in the upstream SOL (scrape-off layer) strongly depend on the divertor operation mode. From the comparison of the W impurity transport between Case A and Case B, obtaining a partially detached state is shown to be effective to reduce W-impurities in the upstream SOL. The limitations of the employed model and the validity of the above results are discussed and future problems are summarized for further applications of IMPGYRO code to ITER plasmas.</p>

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Physics and operation oriented activities in preparation of the JT-60SA tokamak exploitation

Giruzzi G., Yoshida M., Artaud J. F., Asztalos O., Barbato E., Bettini P., Bierwage A., Boboc A., Bolzonella T., Clement-Lorenzo S., Coda S., Cruz N., Day Chr, De Tommasi G., Dibon M., Douai D., Dunai D., Enoeda M., Farina D., Figini L., Fukumoto M., Galazka K., Galdon J., Garcia J., Garcia-Munoz M., Garzotti L., Gil C., Gleason-Gonzalez C., Goodman T., Granucci G., Hayashi N., Hoshino K., Ide S., Imazawa R., Innocente P., Isayama A., Itami K., Joffrin E., Kamada Y., Kamiya K., Kawano Y., Kawashima H., Kobayashi T., Kojima A., Kubo H., Lang P., Lauber Ph, De La Luna E., Maget P., Marchiori G., Mastrostefano S., Matsunaga G., Mattei M., McDonald D. C., Mele A., Miyata Y., Moriyama S., Moro A., Nakano T., Neu R., Nowak S., Orsitto F. P., Pautasso G., Pégourié B., Pigatto L., Pironti A., Platania P., Pokol G. I., Ricci D., Romanelli M., Saarelma S., Sakurai S., Sartori F., Sasao H., Scannapiego M., Shimizu K., Shinohara K., Shiraishi J., Soare S., Sozzi C., Stȩpniewski W., Suzuki T., Suzuki Y., Szepesi T., Takechi M., Tanaka K., Terranova D., Toma M., Urano H., Vega J., Villone F., Vitale V., Wakatsuki T., Wischmeier M., Zagórski R.

Nuclear Fusion 57 （ 8 ） 085001 2017年07月

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

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<p>The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.</p>

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Progress of divertor study on DEMO design

K. Hoshino, N. Asakura, S. Tokunaga, K. Shimizu, Y. Homma, Y. Someya, H. Utoh, Y. Sakamoto, K. Tobita, Joint Special Team for Demo Design

Plasma Fusion Research 12   1405023 2017年05月

研究論文（学術雑誌）, 共著, 筆頭著者, 査読有り

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Conceptual design of imaging bolometer for use of computed tomography in JT-60SA

Sano Ryuichi, Mukai Kiyofumi, Peterson Byron J., Fukumoto Masakatsu, Hoshino Kazuo

Review of Scientific Instruments 88 （ 5 ） 053506 2017年05月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  0034-6748

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<p>An InfraRed imaging Video Bolometer (IRVB) system in JT-60SA has been designed for the two-dimensional (2D) Computed Tomography (CT) analysis of radiation. To achieve complete viewing of the whole poloidal cross section for plasmas having a low aspect ratio, a new IRVB concept, which has two sets of pin-hole camera systems viewing two different directions, has been introduced. Fields of view of the IRVB have been successfully designed by moving apertures to be sufficient for CT use without installation in tangential ports. The size of the aperture, which determines the width of the sightline of the IRVB channels, has also been optimized with respect to the reconstruction error with well-established solvers of tomography. The optimization indicates that the 7 mm square aperture is the best for this system, and the CT measurement without this optimization can become degraded. A synthetic image using an estimated radiation profile shows that the signal to noise (S/N) ratio of the designed IRVB is large enough to identify the incident radiation power. The result indicates that the designed IRVB can be used for the 2D CT measurement of radiation in JT-60SA.</p>

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Estimation of tritium permeation rate to cooling water in fusion DEMO condition

Katayama Kazunari, Someya Youji, Tobita Kenji, Nakamura Hirofumi, Tanigawa Hisashi, Nakamura Makoto, Asakura Nobuyuki, Hoshino Kazuo, Chikada Takumi, Hatano Yuji, Fukada Satoshi

Fusion Science and Technology 71 （ 3 ） 261 - 267 2017年04月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  1536-1055

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<p>The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle &amp; Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium twocomponent system. In the representative recent DEMO condition, the following tritium permeation rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day.</p>

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Simulation of collisional effects on divertor pumping in JT-60SA

Gleason-Gonzalez, C., Varoutis, S., Luo, X., Shimizu, K., Nakano, T., Hoshino, K., Kawashima, H., Asakura, N., Day, Chr., Sakurai, S.

FUSION ENGINEERING AND DESIGN 109   693 - 699 2016年11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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Safety studies of plasma-wall events with AINA code for Japanese DEMO

Rivas, J. C., Nakamura, M., Someya, Y., Hoshino, K., Asakura, N., Takase, H., Miyoshi, Y., Utoh, H., Tobita, K., Dies, J., de Blas, A., Riego, A., Fabbri, M.

FUSION ENGINEERING AND DESIGN 109   1653 - 1657 2016年11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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Simulation of Radiative Divertor Plasmas by Ar Seeding with the Full W-Wall in JT-60SA

Kawashima, H., Shimizu, K., Hoshino, K., Nakano, T., Asakura, N.

CONTRIBUTIONS TO PLASMA PHYSICS 56 （ 6-8 ） 778 - 783 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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Simulation Study of Detached Plasmas by Using One-Dimensional SOL-Divertor Fluid Code with Virtual Divertor Model

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

CONTRIBUTIONS TO PLASMA PHYSICS 56 （ 6-8 ） 729 - 735 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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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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