論文 - 星野　一生

Machine learning-based hydrogen recycling model for predicting rovibrational distributions of released molecular hydrogen on tungsten materials via molecular dynamics simulations

Saito S., Iida M., Nakamura H., Sawada K., Hoshino K., Kobayashi M., Hasuo M., Homma Y., Yamoto S.

Nuclear Materials and Energy 43 2025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

　概要を見る

Understanding the hydrogen recycling process is crucial for comprehending the behavior of detached plasma in nuclear fusion devices. To achieve this, a molecular dynamics (MD) model is being developed to predict the distribution of translational energies and rovibrational states of hydrogen atoms and molecules released from the plasma-facing materials. Neutral transport simulations, utilizing distributions obtained from the MD model as boundary conditions, are also a powerful tool for analyzing the impact of recycled hydrogens on edge plasma. However, the MD model requires significant computational resources to obtain distributions under varying material and irradiation conditions such as material temperature and incident energy. Therefore, developing effective models that seamlessly integrate neutral transport simulation with hydrogen recycling models is crucial. Machine learning techniques are employed to develop predictive models capable of forecasting distributions of energies and rovibrational states of released hydrogen atoms and molecules. Furthermore, a model considering the incident energy distribution (shifted-Maxwellian) is developed by integrating the monochromatic distribution with the shifted-Maxwellian distribution.

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Improvement of a CSD model toward detached plasma analyses by introducing effective loss factors based on plasma spatial distributions

Yamada K., Hoshino K., Hatayama A.

Plasma Physics and Controlled Fusion 67 （ 4 ）  2025年03月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  07413335

　概要を見る

CSD (Core, SOL and Divertor) models sampling a fusion device with three points (Core, SOL and Divertor) aim at understanding fusion plasmas in a wide range of parameter space. Conventional CSD models estimate interactions among plasmas, neutrals and impurities, which are incorporated as loss factors in basic equations in models, with the use of physical quantities of plasmas at three sampled points. Physical quantities are assumed to be uniform in each of the three regions. Neglecting plasma spatial distributions in a divertor prevents the models from accurately evaluating loss factors at low temperatures. This is because loss factors strongly depend on the location-dependent temperature when the temperature is low. Consequently, conventional CSD models have difficulties in reproduction of the detachment. This research increases sampling points for the calculation of loss factors in a divertor. Loss factors at each point are integrated in the volume, then effective loss factors (ELF) for the divertor are obtained. Conventional loss factors in transport equations are replaced with ELFs for an improved CSD model proposed in this study. As a result, the plasma detachment is successfully reproduced in the CSD model without changing the equations. A CSD model for qualitative analyses with low computational costs has been developed to study plasma detachment.

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DISCOVER: New integrated code to simulate helium detached plasma in a linear device

Natsume H., Masuda S., Tanaka H., Migita R., Mano R., Sugiura K., Sawada K., Hoshino K., Hayashi Y., Ohno N.

Physics of Plasmas 32 （ 2 ） 022508 2025年02月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  1070664X

　概要を見る

A detached plasma integrated simulation code with various elastic/inelastic reactions (DISCOVER) was developed for better understanding of detached helium plasma experiments. DISCOVER, which includes a neutral transport code, fluid code, and collisional radiative code, considers elastic and inelastic collisions between recycling neutrals and bulk particles, as well as the transport of metastable atoms [He( 21S) and He( 23S)]. The code was used for simulating helium plasma in the linear plasma device NAGDIS-II. The transition from attached to detached plasma was demonstrated in the simulation by increasing the upstream electron density through an increase in the gas supply from the plasma source while maintaining a fixed downstream gas injection rate. In addition, similar transition was also observed by introducing neutral particles from downstream, while maintaining a fixed upstream electron density. In the detached plasma, the density of He( 23S) near the end target was approximately seven times higher than that in the attached plasma. Furthermore, it was evident that the transport of He( 23S) enhanced the electron energy rate coefficient and influenced the spatial distribution of the detached plasma.

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Investigation of spatial distribution characteristics of emission intensity in detachment on the divertor simulator TPDsheet-ICR

Tonegawa A., Okada N., Miura K., Natume H., Sato K.N., Hoshino K., Hayashi Y., Masuzaki S., Yamoto S.

Nuclear Materials and Energy 41 2024年12月

研究論文（国際会議プロシーディングス）, 共著, 査読有り

　概要を見る

For the first time, we have revealed the spatial structure of a detached plasma with electron–ion recombination (EIR) and molecular-activated recombination (MAR) in response to changes in ion temperature (Ti) and gas pressure. This was achieved using high-density sheet plasma generated by a linear divertor simulator (TPDsheet-ICR). The parallel-plate electrodes were positioned above and below the sheet plasma (∼1019 m−3). We varied the Ti from 3 to 7 eV using ion cyclotron resonance heating and investigated the parameters of the detached plasma near the target in the divergent field region. Plasma emission intensities at Balmer series (Hα and Hγ) and Fulcher band wavelengths (610 ± 10 nm) were measured with a high-speed camera equipped with an Arbaa prism. Additionally, a Langmuir probe measured the electron density and temperature of the plasma. Our results show that as the Ti increases, EIR generated near the plasma periphery disappears, while MAR gradually forms near the plasma center during the transition from detached to attached plasma. This transition occurs because the increase in Ti in the detached plasma transfers energy from ions to electrons, raising the high-energy component of the electrons, resulting in EIR disappearance and the onset of MAR.

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Evaluation of local erosion and deposition on the W-monoblock of JA-DEMO divertor

Oya M., Hoshino K., Asakura N., Sakamoto Y., Ohno N., Hanada K.

Nuclear Materials and Energy 41 2024年12月

研究論文（国際会議プロシーディングス）, 共著, 査読有り

　概要を見る

We developed and performed a 3D Monte-Carlo simulation on local erosion and deposition in outer divertor target of JA-DEMO. The local erosion and deposition were evaluated considering the divertor plasma parameters and the roof shape of W-monoblocks. First, for local erosion simulation, the trajectories of D ions and impurity (He and Ar) ions were traced before irradiating the monoblock surface. The physical sputtering was calculated by these ions irradiation, considering their charge states, incident-energies and -angles. The highest sputtering flux was produced by multiply-charged Ar ions (Ar3+–Ar5+). The sputtering flux strongly depended on the poloidal positions along the divertor target. Although almost no erosion (sputtering) was obtained in low-temperature zone (∼1 eV), the erosion occurred in mid- (∼10 eV) and high-temperature (>10 eV) zone. Second, for local deposition simulation, the trajectories of W atoms emitted from the monoblock surface were traced. After transport (including ionization and recombination) in the plasma, W atoms and ions were deposited on the monoblock surface. The deposition occurred even on the area where plasmas did not irradiate (shadow). This was due to perpendicular diffusion to magnetic field lines, gyro-orbit effects for W ions and straight-line motion for neutral W atoms. The poloidal distribution of the deposition fluxes was similar to that of the sputtering fluxes. In low-temperature zone, no deposition flux occurred due to very low sputtering flux. The deposition flux in mid-temperature zone increased significantly and that in high-temperature zone remained high.

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Analyses of Particle and Energy Balances using a Core-SOL-Divertor (CSD) Model for Tokamak DEMO-Reactor Studies

Yamada K., Hoshino K., Hatayama A.

Plasma and Fusion Research （一般社団法人 プラズマ・核融合学会）  19 （ 0 ） 1403034 - 1403034 2024年11月

研究論文（学術雑誌）, 共著, 査読有り

　概要を見る

Understanding of key physics and engineering parameters is one of the most important issues to conduct fundamental design studies of future fusion reactors. Also, sensitivity analyses of performance by parameter surveys need to be done for optimizing a reactor. This research aims at a qualitative analysis to identify directions of design parameter optimization for fusion reactors. For this purpose, a Core-SOL-Divertor (CSD) model has been employed, because it is suitable for a wide range of parameter surveys by changing various parameters with low computational costs. This paper analyzes the particle and energy balances of the JT-60U divertor plasma by using a CSD model and reveals dependencies of the balances on design parameters of the device. In addition, the same CSD model is applied to conduct divertor plasma analyses for a future demonstration reactor (JA-DEMO). The result shows that it is possible to obtain a low temperature state of its divertor plasma as has been already reported by a two-dimensional SOL-Divertor integrated code. From these results, this paper shows that the CSD model is applicable to basic studies of directions and design concepts for future fusion reactors.

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Recent progress of JT-60SA project toward plasma operation

Shirai H., Takahashi K., Di Pietro E., Abate D., Abdel Maksoud W., Abe H., Aiba N., Abe T., Akimitsu M., Ayllon-Guerola J., Arai T., Artaud J.F., Asakura N., Ashikawa N., Balbinot L., Barabaschi P., Baulaigue O., Belonohy E., Belpane A., Bin W., Bombarda F., Bolzonella T., Bonne F., Bonotto M., Botija J., Buermans J., Cabrera-Pérez S., Cardella A., Carralero D., Carraro L., Cavalier J., Cavinato M., Chernyshova M., Chiba S., Clement-Lorenzo S., Cocilovo V., Coda S., Coelho R., Coffey I., Collin B., Corato V., Cucchiaro A., Czarski T., Dairaku M., Davis S., Day C., Dela Luna E., De Tommasi G., Decool P., Di Pace L., Dibon M., Disset G., D’Lsa F., Ejiri A., Endo Y., Ezumi N., Falchetto G., Fassina A., Fejoz P., Ferro A., Fietz W., Figini L., Fornal T., Frello G., Fujita T., Fukuda T., Fukui K., Fukumoto M., Funaba H., Furukawa M., Futatani S., Gabellieri L., Gaio E., Galazka K., Garcia J., Garcia-Dominguez J., Garcia-Lopez J., Garcia-Munoz M., Garzotti L., Gasparini F., Gharafi S., Giacomelli L., Ginoulhiac G., Giruzzi G., Giudicotti L., Gonzalez-Martin J., Guillén-González R., Hajnal N., Hall S., Hamada K., Hanada K., Hanada M., Hasegawa K., Hatakeyama S., Hauer V., Hayashi N., Hayashi T., Heller R., Hidalgo-Salaverri J., Higashijima S.

Nuclear Fusion 64 （ 11 ）  2024年11月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00295515

　概要を見る

Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

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Emission of high rovibrational hydrogen molecules under detached plasma conditions by recycling on the tungsten wall

Saito S., Nakamura H., Sawada K., Hoshino K., Kojima Y., Doi T., Kobayashi M., Hasuo M., Homma Y., Yamoto S.

Nuclear Fusion 64 （ 12 ） 126067 2024年10月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00295515

　概要を見る

It is well known that the rate coefficient of molecular-assisted recombination (MAR) varies by several orders of magnitude depending on the rovibrational states of the hydrogen molecules. A molecular dynamics simulation is performed to estimate the rovibrational states of recycled hydrogen molecules at the divertor in the JA-DEMO reactor under detached plasma conditions. The simulation results reveal that molecules in high rovibrational states are released even with low incident energy, which will be the dominant condition under detached plasma conditions. Molecules generated in this way can strongly affect the formation of the detached plasma via a molecular assisted-process such as MAR.

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Deep learning model for predicting the spatial distribution of binding energy from atomic configurations

Saito S., Sato S., Nakamura H., Takahashi C., Sawada K., Hoshino K., Kobayashi M., Hasuo M.

Japanese Journal of Applied Physics 63 （ 9 ）  2024年08月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00214922

　概要を見る

Understanding plasma-material interaction is crucial for achieving steady-state operation of magnetic confinement fusion devices. Kinetic Monte Carlo (kMC) simulation is a powerful tool for investigating the motion of atoms in the plasma facing materials under the influence of this interaction. To predict trapping sites and migration energies necessary for kMC simulations, we developed a deep learning model based on pix2pix for predicting the spatial distribution of binding energy. Results show that the model can reproduce spatial distributions similar to the true values. However, larger errors occur in regions with steep value gradients.

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Numerical study of the plasma meniscus shape and beam optics in RF negative ion sources

Hayashi K., Hoshino K., Miyamoto K., Hatayama A., Lettry J.

Journal of Instrumentation 19 （ 4 ） C04031 2024年04月

研究論文（国際会議プロシーディングス）, 共著, 査読有り

　概要を見る

In order to extract intense ion beams with good beam optics from hydrogen negative ion sources, it is important to control the shape of the plasma meniscus (i.e. beam emission surface). Recently, it is suggested experimentally that the degradation of beam optics in the RF negative ion sources may be due to the fluctuation of the distance d eff between the meniscus and the extraction grid caused by the fluctuation of the plasma density np . The purpose of this study is to make clear the dependence of d eff on np in the presence of a large amount of surface produced H- ions in order to understand such fluctuation of beam optics in RF sources For the purpose, 3D electrostatic PIC simulation was conducted taking the bulk plasma density as a parameter, investigating the extraction region of a H- ion source. A large amount of the surface H- production on the PG has been taken into account under the assumption that the H- production rate is proportional to the bulk plasma density. The result shows that the effective distance d eff is proportional to np-1/2 even for a large amount of surface H- production. This dependence suggests that the bulk plasma density np is the key parameters to control d eff and the resultant beam optics extracted from the negative ion source.

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Hamiltonian simulation using the quantum singular-value transformation: Complexity analysis and application to the linearized Vlasov-Poisson equation

Toyoizumi K., Yamamoto N., Hoshino K.

Physical Review A 109 （ 1 ） 01243 2024年01月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  24699926

　概要を見る

Quantum computing can be used to speed up the simulation time (more precisely, the number of queries of the algorithm) for physical systems; one such promising approach is the Hamiltonian simulation (HS) algorithm. Recently, it was proven that the quantum singular-value transformation (QSVT) achieves the minimum simulation time for HS. An important subroutine of the QSVT-based HS algorithm is the amplitude amplification operation, which can be realized via the oblivious amplitude amplification or the fixed-point amplitude amplification in the QSVT framework. In this work we execute a detailed analysis of the error and number of queries of the QSVT-based HS and show that the oblivious method is better than the fixed-point one in the sense of simulation time. Based on this finding, we apply the QSVT-based HS to the one-dimensional linearized Vlasov-Poisson equation and demonstrate that the linear Landau damping can be successfully simulated.

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Spatial Distribution Analyses of Axially Long Plasmas under a Multi-Cusp Magnetic Field Using a Kinetic Particle Simulation Code KEIO-MARC

Nishimura R., Seino T., Yoshimura K., Takahashi H., Matsuyama A., Hoshino K., Oishi T., Tobita K.

Plasma （Plasma）  7 （ 1 ） 64 - 75 2024年01月

研究論文（学術雑誌）, 共著, 査読有り

　概要を見る

To realize the development of a long plasma source with a uniform electron density distribution in the axial direction, the spatial distribution of plasma under a multi-cusp magnetic field was analyzed using a KEIO-MARC code. Considering a cylindrical plasma source with an axial length of 3000 mm and a cross-sectional diameter of 100 mm, in which the filament electrode was the electron source, the electron density distribution was calculated using the residual magnetic flux density, Bres, and the number of permanent magnets installed at different locations surrounding the device, Nmag, as design parameters. The results show that both Bres and Nmag improved the uniformity of the electron density distribution in the axial direction. The maximum axial electron density decreased with increasing Nmag and increased with increasing Bres. These trends can be explained by considering the nature of the multi-cusp field, where particles are mainly confined to the field-free region (FFR) near the center of the plasma column, and the loss of particles due to radial particle transport. The use of multiple filaments at intervals shorter than the plasma decay length dramatically improved axial uniformity. To further improve axial uniformity, the filament length and FFR must be properly set so that electrons are emitted inside the FFR.

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Effect of Large-Angle Scattering on Particle Transport and Heat Flux in Advanced Divertor Magnetic Field Configurations

Umezaki D., Matsuura H., Hoshino K.

Plasma and Fusion Research （一般社団法人 プラズマ・核融合学会）  19 （ 0 ） 1403003 - 1403003 2024年01月

研究論文（学術雑誌）, 共著, 査読有り

　概要を見る

Divertor heat loads are one of the most significant issues affecting fusion reactors. Atomic processes play a crucial role in reduction of the divertor heat load. Notably, elastic scattering between ions and neutral particles can be characterized as large-angle scattering. A large fraction of ion energy is transferred to neutral particles, and the ion direction can be significantly changed by a single large-angle scattering event. In abundant neutral particle regions such as divertor plasmas, the large-angle elastic scattering results in additional ion transport perpendicular to magnetic field lines. Effect of the additional ion transport is expected to be significant at low magnetic field strength and long Larmor radii, such as in a case of advanced divertors (e.g., Super-X and Snowflake divertors). In this study, we investigated the effect of the large-angle elastic scattering at low magnetic field strength and long divertor legs with reference to advanced divertor configurations using an orbital calculation. The large-angle elastic scattering transport is seen to cause a spread in density profiles and a reduction of heat flux. The results of this study show that for the short (long) leg divertor configuration like JT-60U (advanced divertor), the peak heat flux is reduced by around 15% (21%) when the magnetic field strength is 0.5 T in comparison to the model that assumes no guiding center movement due to the elastic scattering. It is also shown that the assumption of isotropic elastic scattering with neutral particles leads to excessive suppression of ion flows.

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SOL impurity transport analysis with SONIC code with a kinetic effect on a thermal force transport model in JT-60U

Kobayashi M., Yamoto S., Hoshino K., Asakura N., Homma Y., Aiba N., Nakano T., Sano R., Kawamura G., Hayashi Y.

Contributions to Plasma Physics （Contributions to Plasma Physics）  64 （ 7-8 ）  2024年

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  08631042

　概要を見る

The latest version of the SONIC code with the extended thermal force model that is capable to treat collisionality dependence of the force has been applied to the interpretative simulation of L-mode discharge of JT-60 U to investigate the impact of kinetic correction of the thermal force on the scrape-off layer (SOL) plasma properties. As a first attempt, we have selected an L-mode attached discharge #39090 (Ip = 1.6 MA, Bt = 3.1 T, and PNBI = 4.5 MW), where (Formula presented.) = 2.08 × 1019 m−3, (Formula presented.) = 0.392 with deuterium plasmas. It has been found that the peak values of SOL impurity density are reduced by a factor of 2, with the kinetic correction, because the thermal force is reduced by the low collisionality of SOL plasma between the X-point and the (inner and outer) midplane. On the other hand, the profiles near the divertor plates are unaffected by the kinetic correction due to the highly collisional condition. The kinetic correction is found more pronounced for the higher charge states impurity due to the charge dependence of the parallel impurity force balance. Overall effects due to the kinetic correction on the total impurity radiation and the divertor heat load are confirmed to be a few % in the present discharge condition of JT-60U.

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Estimation of impact of non-local ion heat flux model on transport simulation for DEMO-relevant scrape-off layer plasma

Homma Y., Hoshino K., Yamoto S., Tokunaga S., Asakura N., Miyato N.

Contributions to Plasma Physics 64 （ 7-8 ）  2024年

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  08631042

　概要を見る

In a fusion DEMOnstration reactor (DEMO)-relevant scrape-off layer plasma (SOL), whose collisionality is lower than in the SOLs of present-day tokamaks, kinetic effects are predicted to reduce the plasma heat conductivity along the magnetic field below the value obtained with the classical Spitzer–Harm model. As a part of ongoing efforts to improve the predictive capability of SOL heat transport calculations, we have implemented the non-local heat flux model proposed by [Luciani, Mora and Virmont, Phys. Rev. Lett. 51 (1983) 1664–1667] (here referred to as the LMV model) in the integrated SOL–divertor simulation code SONIC in order to account for the kinetic effect on the heat conduction due to the parallel streaming of rapidly moving particles. Our transport simulations for the Japanese demonstration tokamak reactor concept, JA DEMO, show that the LMV model yields a significantly reduced ion parallel conductive heat flux density both on the low- and high-field side of the upstream SOL. The heat flux obtained with the LMV model seems to be consistent with earlier kinetic simulations of tokamak ion transport. As a consequence of the reduced heat flux, a significant increase in the ion temperature and a decrease in the density have also been found over a broad area of the upstream SOL.

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Simple 3D PIC Analysis for Beam Phase Space Oscillation in RF Driven Negative Hydrogen Ion Source

Shibata T., Shinto K., Nakano H., Hoshino K., Miyamoto K., Ohkoshi K., Nanmo K., Ikegami K., Kawai I., Oguri H., Wada M.

Journal of Physics: Conference Series （Journal of Physics: Conference Series）  2743 （ 1 ） 012007 2024年

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  17426588

　概要を見る

Temporal oscillation of the negative hydrogen ion (H-) beam in the Radio Frequency (RF) ion source is investigated by a simple 3D3V Particle-In-Cell (PIC) model. The model solves electron, proton, and H- transport processes in the extraction region. The calculation domain is from the vicinity of the beam aperture to the ground electrode. To understand the relation between the plasma density oscillation and the extracted H- beam characteristics, the input electron and proton fluxes from the driver region are varied parametrically with the fundamental J-PARC RF frequency (2 MHz). The numerical results give an idea of the main physical processes between the oscillations of the plasma parameters and the extracted H- ion trajectories depending on the plasma meniscus shape in the different RF phases.

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First results of multi-fluid modeling of detached hydrogen plasmas in a linear plasma device using fluid code LINDA-NU

Sugiura K., Ido T., Tanaka H., Natsume H., Masuda S., Hoshino K., Sawada K., Ohno N.

Contributions to Plasma Physics （Contributions to Plasma Physics）  64 （ 7-8 ）  2024年

研究論文（国際会議プロシーディングス）, 共著, 査読有り,  ISSN  08631042

　概要を見る

In order to simulate hydrogen (H) plasma in the linear plasma device NAGDIS-II, we have modified the fluid code LINDA-NU to allow the simultaneous calculation of multiple ion species consisting of hydrogen atomic ions ((Formula presented.)) and molecular ions ((Formula presented.)). In this simulation, H and (Formula presented.) neutrals are assumed to be uniformly distributed in space in order to obtain initial qualitative results. The fraction of (Formula presented.) ions increases as the (Formula presented.) molecular density increases, and the recombination process between (Formula presented.) and electrons is observed to reduce the particle flux to the target plate. With an increase in H density, the electron density increases due to the decrease in ion flow velocity due to the change exchange process, and the electron temperature decreases to less than 1 eV, leading to the detached plasma formation attributed to the electron-ion recombination process.

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Effect of elastic scattering between ions and neutral particles on ion transport perpendicular to magnetic fields in divertor plasmas

Umezaki D., Matsuura H., Hoshino K.

Contributions to Plasma Physics 64 （ 1 ） e202300064 2023年09月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  08631042

　概要を見る

In divertor plasmas, atomic processes play a significant role in reducing the divertor heat load. In particular, the elastic scattering between ions and neutral particles can be characterized as a large-angle scattering; this is in contrast with Coulomb scattering, which is dominated by small-angle scattering. In a large-angle scattering, a large fraction of the ion energy is transferred to the neutral particle, and the particle flight direction can be changed significantly. This process can generate additional particle transport perpendicular to magnetic field lines and affect plasma density profile in neutral-rich divertor regions. However, in most edge plasma simulation codes, the elastic scattering with neutral particles is not taken into account in the direction perpendicular to the magnetic field lines. In this study, we performed a 2D orbital calculation without other drifts such as diamagnetic and E × B drifts to assess the effect of the elastic scattering on the density profile. The peak density is found to be decreased by 11.1 (6.3)% compared with the model that ignores the transport due to the elastic scattering, even at 3.0 (5.0) T magnetic field similar to that of the JT-60SA (JA DEMO reactor) class. In addition, the ion transport can be expressed as a diffusion model by integrating the differential cross section over the large-angle scattering range (Formula presented.). The proposed method may be easily introduced into any integrated codes to consider the elastic scattering.

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The development of a zero-dimensional collisional-radiative model for interpreting plasma emission in low temperature divertor plasmas in tokamaks

Fox-Widdows E., Bowden M.D., Hoshino K., Hatayama A., Osawa R., Tsubotani Y.

Frontiers in Physics 11   1154185 2023年07月

研究論文（学術雑誌）, 共著, 査読有り

　概要を見る

Collisional-radiative models are commonly used to analyse atomic and molecular processes in low temperature plasmas by determining the distribution functions of excited states as functions of various plasma parameters. This paper outlines the improvements to a zero-dimensional collisional-radiative model, developed at Keio University, for purposes relevant to the analysis of emission measurements in low temperature hydrogen plasmas. The extension of this 0D model centres on the inclusion of additional molecular species and reactions, a calculation of emission intensity in order to directly compare with experimental work, and the addition of a simple wall model to allow for a deeper understanding of recycling of atoms and molecules in low temperature plasmas close to plasma facing components in fusion machines. Results from the improved model are then compared with both an existing CR model for benchmarking and experimental emission spectroscopy data from an inductively-coupled plasma device at the University of Liverpool. These results show how the developments to the model have increased the relevance to experimental plasmas, such as those in the power exhaust regions of fusion machines, with the ratio of the Hβ/Hα line intensities outputted from the improved CR model mirroring more closely the emission measured experimentally in the ICP device.

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Transient analysis of high-Z impurity screening by additional injection of low-Z impurity using integrated divertor code SONIC

Yamoto S., Hoshino K., Homma Y., Nakano T., Hayashi N.

Nuclear Fusion （Nuclear Fusion）  63 （ 7 ） 076019 2023年06月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00295515

　概要を見る

The dynamics of the screening effect of Ar impurity by the injection of additional Ne has been studied through time-dependent analysis with the integrated divertor code SONIC. In the preceding study (Yamoto et al 2020 Plasma Phys. Control. Fusion 62 045006), the predictive simulation of JT-60SA plasma by SONIC has shown that the injection of additional Ne into Ar-seeded plasma results in lower Ar density and radiation power in the SOL and core edge than in the Ar-only seeded case. The results have demonstrated that the mixed impurity seeding of Ar and Ne may be advantageous for maintaining a high core plasma performance with a low divertor heat load. It was found that the friction force induced by the high D+ flow in the SOL towards the inner divertor (ID) region in the Ar + Ne seeded case pushes Ar impurities to the ID. However, the dynamics of D+ flow acceleration cannot be interpreted in the previous study because SONIC was a steady state code. In this study, we have developed the time-dependent version of SONIC and applied it to the transient analysis of the injection of additional Ne into Ar-seeded plasma in JT-60SA. When additional Ne is injected, Ne ions stay in the ID plasma near the X-point. As a result, the Ne radiation power increases near the X-point. The electron pressure then decreases due to the radiation cooling and the D+ flow is accelerated by the electron pressure gradient. The ion pressure also decreases due to the convection by the accelerated D+ flow by electron pressure gradient. The resulting ion pressure gradient further accelerates the D+ flow velocity towards the ID. The results suggest that both the high-performance core plasma and the low divertor heat load can be achieved by the Ar + Ne mixed impurity seeding.

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Study of plasma meniscus including surface produce negative ions by using PIC-MCC simulation

Miyamoto K., Hayashi K., Hoshino K., Hatayama A.

Journal of Instrumentation 18 （ 6 ） C06014 2023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

　概要を見る

In general, a quality of negative ion beam optics is determined by the shape of the plasma meniscus, which is an ion emitting surface. In this study, the key parameters to control the plasma meniscus and relevant physical structure in the electronegative plasma including the surface produced H- ions is investigated by using PIC-MCC simulation. The region from the source plasma up to the accelerator is modeled for a single aperture, and thus, the plasma meniscus can be obtained self-consistently. Parameter survey for the net flux of surface produced H- ion is conducted by varing the numbers of the surface produced H- ion super-particles per timestep in order to investigate the effcet of the electronegativity, that is, the ratio of the negative ion density to the electron density. It is shown that the effective distance d eff, between the plasma meniscus and the extraction grid depends on the electronegativity as well as the plasma density. Especially, the effective distance d eff, decreases with the increase of the electronegativity under the constant plasma density, which means that the shape of the plasma meniscus becomes flat or convex rather than concave. Moreover, it is also verified that the electronegativity affects the H- ion trajectories extracted from near the edge of plasma meniscus through the shape of plasma meniscus.

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Recent progress of plasma exhaust concepts and divertor designs for tokamak DEMO reactors

Asakura N., Hoshino K., Kakudate S., Subba F., You J.H., Wiesen S., Rognlien T.D., Ding R., Kwon S.

Nuclear Materials and Energy 35   101446 2023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

　概要を見る

The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2. Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

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Effect of Surface Produced H̄ Ion on the Plasma Meniscus in Negative Hydrogen Ion Sources

Hayashi K., Hoshino K., Hatayama A., Miyamoto K., Lettry J.

Plasma and Fusion Research （一般社団法人 プラズマ・核融合学会）  18 （ 0 ） 1401008 - 1401008 2023年03月

研究論文（学術雑誌）, 共著, 査読有り

　概要を見る

To extract intense ion beams with good beam optics from ion sources, controlling the distance deff between the plasma meniscus (i.e., beam emission surface) and the beam extraction grid is important. This study conducts a novel investigation into the dependence of the effective distance deff on the amount of surface H̄ production SH-. For this purpose, a 3D PIC (three dimensional Particle-in-Cell) simulation is conducted to obtain a model geometry of the extraction region for a H̄ ion source with SH- as a parameter. Based on results, deff significantly depends on SH- and the H̄-electron density ratio (α = nH̄=ne) in front of the extraction aperture for the same plasma density; as SH- increases, deff decreases. The results suggest that SH- is critical for controlling deff and the resultant beam optics extracted from the negative ion source

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Analysis of plasma meniscus in a hydrogen electronegative plasma

Miyamoto K., Hayashi K., Hoshino K., Hatayama A.

Plasma Sources Science and Technology 31 （ 10 ） 105012 2022年11月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  09630252

　概要を見る

The emitting surface of an ion beam (plasma meniscus) in the vicinity of the extraction aperture is theoretically investigated. In the present study, the electrons, H+ ions and volume produced H− ions are contained in the electronegative plasma. By solving the one-dimensional Poisson equation, an analytical formula is obtained from the balance between the current reaching a wall and the space charge limited current of negatively charged particles. The analytical formula indicates that the distance, d eff, between the plasma meniscus and the extraction grid includes the ratio of masses of the negative ions and electrons, which is different from that for a plasma containing electrons and single positive ions, and depends on the ratio of the negative ion density to the electron density, α. With increase in the value of α, the distance d eff becomes shorter, and thus the penetration of the electric field for negative ion extraction into the source plasma becomes smaller. This tendency is confirmed by using a three-dimensional particle-in-cell simulation from the results of the contour map of the electrostatic potential near the plasma grid and the location of the plasma meniscus. The reasons why the distance d eff becomes shorter with increase in the value of α are considered to be the larger space charge effect of the negative ions compared with that of the electrons and the decrease in the plasma potential.

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Plasma physics and control studies planned in JT-60SA for ITER and DEMO operations and risk mitigation

Yoshida M., Giruzzi G., Aiba N., Artaud J.F., Ayllon-Guerola J., Balbinot L., Beeke O., Belonohy E., Bettini P., Bin W., Bierwage A., Bolzonella T., Bonotto M., Boulbe C., Buermans J., Chernyshova M., Coda S., Coelho R., Davis S., Day C., De Tommasi G., Dibon M., Ejiri A., Falchetto G., Fassina A., Faugeras B., Figini L., Fukumoto M., Futatani S., Galazka K., Garcia J., Garcia-Muñoz M., Garzotti L., Giacomelli L., Giudicotti L., Hall S., Hayashi N., Hoa C., Honda M., Hoshino K., Iafrati M., Iantchenko A., Ide S., Iio S., Imazawa R., Inoue S., Isayama A., Joffrin E., Kamiya K., Ko Y., Kobayashi M., Kobayashi T., Kocsis G., Kovacsik A., Kurki-Suonio T., Lacroix B., Lang P., Lauber P., Louzguiti A., De La Luna E., Marchiori G., Mattei M., Matsuyama A., Mazzi S., Mele A., Michel F., Miyata Y., Morales J., Moreau P., Moro A., Nakano T., Nakata M., Narita E., Neu R., Nicollet S., Nocente M., Nowak S., Orsitto F.P., Ostuni V., Ohtani Y., Oyama N., Pasqualotto R., Pégourié B., Perelli E., Pigatto L., Piccinni C., Pironti A., Platania P., Ploeckl B., Ricci D., Roussel P., Rubino G., Sano R., Särkimäki K., Shinohara K., Soare S., Sozzi C., Sumida S., Suzuki T., Suzuki Y.

Plasma Physics and Controlled Fusion 64 （ 5 ） 054004 2022年05月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  07413335

　概要を見る

A large superconducting machine, JT-60SA has been constructed to provide major contributions to the ITER program and DEMO design. For the success of the ITER project and fusion reactor, understanding and development of plasma controllability in ITER and DEMO relevant higher beta regimes are essential. JT-60SA has focused the program on the plasma controllability for scenario development and risk mitigation in ITER as well as on investigating DEMO relevant regimes. This paper summarizes the high research priorities and strategy for the JT-60SA project. Recent works on simulation studies to prepare the plasma physics and control experiments are presented, such as plasma breakdown and equilibrium controls, hybrid and steady-state scenario development, and risk mitigation techniques. Contributions of JT-60SA to ITER and DEMO have been clarified through those studies.

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Development and Application of SONIC Divertor Simulation Code to Power Exhaust Design of Japanese DEMO Divertor

Asakura N., Hoshino K., Homma Y., Sakamoto Y.

Processes （Processes）  10 （ 5 ） 872 2022年04月

研究論文（学術雑誌）, 共著, 査読有り

　概要を見る

An integrated divertor simulation code, SONIC, An integrated divertor simulation code, SONIC, has been developed in order to predict a self-consistent transport solution of the plasma, neutral and impurities in the scrape-off layer (SOL) and divertor. SONIC code has contributed to determining the divertor design and power handling scenarios for the Japanese (JA) fusion demonstration (DEMO) reactor. Radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated to evaluate the power exhaust scenarios with . The simulation identified the decay length of the total parallel heat flux profile as being broader than the electron one, because of the ion convective transport from the outer divertor to the upstream SOL, produced by the plasma flow reversal. The flow reversal also reduced the impurity retention in the outer divertor, which may produce the partial detachment. Divertor operation margin of key power exhaust parameters to satisfy the peak was determined in the low of under severe conditions such as reducing radiation loss fraction, i.e., and diffusion coefficients and . The divertor geometry and reference parameters rad) were consistent with the low operation of the JA DEMO concepts. For either severe assumption of or and to their half values, higher operation was required. In addition, recent investigations of physics models (temperature-gradient force on impurity, photon transport, neutral-neutral collision) under the DEMO relevant SOL and divertor condition are presented.

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Study of negative ion beam emittance characteristic using 3D PIC-MCC simulation

Miyamoto K., Nagaoka K., Hatayama A., Hoshino K., Nakano H., Shibata T., Tsumori K.

Journal of Physics: Conference Series 2244 （ 1 ） 012040 - 012040 2022年04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  17426588

　概要を見る

It is reported from the measurement of a negative ion beamlet emittance that the characteristic three-Gaussian components are observed in the emittance diagram. The negative ion trajectories are calculated using 3D PIC-MCC simulation in order to clarify the origin of this complicated phase space structure and underlying physical mechanism. The characteristic three-Gaussian components can be reproduced in the emittance diagram from the simulation result. It is verified that the Gaussian components are caused by the negative ions extracted from the different plasma meniscus region, that is, the central region or the region near the edges of the meniscus.

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Numerical analysis of isotope effect in NIFS negative ion source

Kato R., Hoshino K., Nakano H., Shibata T., Miyamoto K., Iwanaka K., Hayashi K., Hatayama A.

Journal of Physics: Conference Series 2244 （ 1 ） 012035 - 012035 2022年04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  17426588

　概要を見る

Result of hydrogen (H) and deuterium (D) experiments done by NIFS research and development negative ion source (RNIS) demonstrated that the co-extracted electron current with the negative ions and the electron density in the driver region in the D experiment have been around three times higher than that in the H experiment. To investigate mechanism of this difference, electron transport simulation using 3D kinetic particle tracking model KEIO-MARC code has been modified and applied to analysis of the isotope effect in the NIFS negative ion source. Simulation result suggests that impacts of isotope effects of sheath potential drop, coulomb collisions, and some reactions of ground state molecules and ions on the electron density is not large to explain the experimental result of the increase in the electron density in the plasma.

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Numerical study of different electron density observed in Hydrogen and Deuterium ion source plasmas

Shibata T., Kato R., Nakano H., Hayashi K., Sato J., Miyamoto K., Hoshino K., Hatayama A.

Journal of Physics: Conference Series 2244 （ 1 ） 012002 - 012002 2022年04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  17426588

　概要を見る

Sequences of hydrogen (H) and deuterium (D) experiments have been done by NIFS research and development negative ion source (RNIS) for the deuterium NBI development. In the experiments, the co-extracted electron current with the negative ions and the electron density in the plasma generation region in the D experiment have been around three times higher than that in the H experiment. To explain the difference of the electron density in the RNIS driver region, a zero-dimensional numerical model is developed in the present study. The model only focuses on the isotope effect for vibrationally excited level of electronically grounded state molecules and its relevant cross-sections. The calculation results show that difference of the ionization channel numbers via molecular vibrationally excited states could be a reason to enhance ionization rate in D plasma.

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Completion of JT-60SA construction and contribution to ITER

Kamada Y., Di Pietro E., Hanada M., Barabaschi P., Ide S., Davis S., Yoshida M., Giruzzi G., Sozzi C., Abdel Maksoud W., Abe H., Aiba N., Akiyama T., Ayllon-Guerola J., Arai T., Artaud J.F., Asakura N., Ashikawa N., Balbinot L., Bando T., Baulaigue O., Belonohy E., Bin W., Bombarda F., Bolzonella T., Bonne F., Bonotto M., Botija J., Cabrera-Pérez S., Cardella A., Carraro L., Cavalier J., Chernyshova M., Chiba S., Clement-Lorenzo S., Cocilovo V., Coda S., Coelho R., Coffey I., Collin B., Corato V., Cucchiaro A., Czarski T., Dairaku M., Davis S., Day C., de la Luna E., De Tommasi G., Decool P., Di Pace L., Di Pietro E., Dibon M., Disset G., Ejiri A., Endo Y., Ezumi N., Falchetto G., Fassina A., Fejoz P., Ferro A., Fietz W., Figini L., Fornal T., Frello G., Fujita T., Fukuda T., Fukui K., Fukumoto M., Furukawa M., Futatani S., Gabellieri L., Gaio E., Galazka K., Garcia J., Garcia-Dominguez J., Garcia-Lopez J., Garcia-Munoz M., Garzotti L., Gasparini F., Gharafi S., Giacomelli L., Ginoulhiac G., Giruzzi G., Giudicotti L., Guillén González R., Hajnal N., Hall S., Hamada K., Hanada K., Hanada M., Hasegawa K., Hatae T., Hatakeyama S., Hauer V., Hayashi N., Hayashi T., Heller R., Higashijima S., Hinata J.

Nuclear Fusion 62 （ 4 ） 042002 2022年03月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00295515

　概要を見る

Construction of the JT-60SA tokamak was completed on schedule in March 2020. Manufacture and assembly of all the main tokamak components satisfied technical requirements, including dimensional accuracy and functional performances. Development of the plasma heating systems and diagnostics have also progressed, including the demonstration of the favourable electron cyclotron range of frequency (ECRF) transmission at multiple frequencies and the achievement of long sustainment of a high-energy intense negative ion beam. Development of all the tokamak operation control systems has been completed, together with an improved plasma equilibrium control scheme suitable for superconducting tokamaks including ITER. For preparation of the tokamak operation, plasma discharge scenarios have been established using this advanced equilibrium controller. Individual commissioning of the cryogenic system and the power supply system confirmed that these systems satisfy design requirements including operational schemes contributing directly to ITER, such as active control of heat load fluctuation of the cryoplant, which is essential for dynamic operation in superconducting tokamaks. The integrated commissioning (IC) is started by vacuum pumping of the vacuum vessel and cryostat, and then moved to cool-down of the tokamak and coil excitation tests. Transition to the super-conducting state was confirmed for all the TF, EF and CS coils. The TF coil current successfully reached 25.7 kA, which is the nominal operating current of the TF coil. For this nominal toroidal field of 2.25 T, ECRF was applied and an ECRF plasma was created. The IC was, however, suspended by an incident of over current of one of the superconducting equilibrium field coil and He leakage caused by insufficient voltage holding capability at a terminal joint of the coil. The unique importance of JT-60SA for H-mode and high-β steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, IC and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.

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Impact of neutral gas puffing on the divertor power exhaust and particle control in GAMMA 10/PDX by the LINDA-KNMC code

Islam M.S., Nakashima Y., Hatayama A., Hoshino K., Sakamoto M.

Contributions to Plasma Physics 62 （ 5-6 ） e202100141 2021年12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  08631042

　概要を見る

This paper numerically investigates the impact of neutral gas puffing on the divertor power exhaust and particle control in GAMMA 10/PDX applying the plasma fluid code LINDA and the kinetic neutral code KMNC. The LINDA code is applied to GAMMA 10/PDX plasmas for understanding the physics of detachment and energy loss processes. The LINDA code is coupled with a Monte–Carlo neutral code for describing the hydrogen (H and H2) neutral particles. The plasma ion temperature (Ti) is reduced towards the target plate with the increment of H2 Gas-Puff rate. This paper also numerically investigates the impact of recombination processes on the detached formation. The ionization, charge–exchange, and recombination of H have been included as a particle and an energy sink and source parts of the fluid equations. The reaction rate coefficient of molecular activated recombination (MAR) has been recently introduced in the LINDA code. In this paper, we study numerically the effect of MAR on the detached plasma. The MAR plays a dominant role in making the detached plasma. The simulation clarified that H neutral particles injection can effectively generate the detached plasma by enhancing the plasma-neutral collisions processes.

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Power exhaust concepts and divertor designs for Japanese and European DEMO fusion reactors

Nobuyuki Asakura, Kazuo Hoshino, Satoshi Kakudate, Fabio Subba, Christian Vorpahl, Yuki Homma, Hiroyasu Utoh, Youji Someya, Yoshiteru Sakamoto, Ryoji Hiwatari, Satoshi Suzuki, Jeong-Ha You, Mattia Siccinio, Gianfranco Federici

Nuclear Fusion （{IOP} Publishing）  61 （ 12 ） 126057 2021年10月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00295515

　概要を見る

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan-Europe broader approach DEMO design activity (BA DDA). Common critical issues are the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling (P radtot/P heat 0.8). Different exhaust concepts in the main plasma and divertor have been developed for Japanese (JA) and European (EU) DEMOs. JA proposed a conventional closed divertor geometry to challenge large P sep/R p handling of 30-35 MW m-1 in order to maintain the radiation fraction in the main plasma at the ITER-level (f radmain = P radmain/P heat ∼ 0.4) and higher plasma performance. EU challenged both increasing f radmain to ∼0.65 and handling the ITER-level P sep/R p in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding P sep = 250-300 MW and 150-200 MW, respectively. Both results showed that large divertor radiation fraction (P raddiv/P sep 0.8) was required to reduce both peak q target (10 MW m-2) and T e,idiv. In addition, the JA divertor performance with EU-reference P sep of 150 MW showed benefit of the closed geometry to reduce the peak q target and T e,idiv near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130 °C-200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs.

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Study of numerical error of a Eulerian-Lagrangian scheme in the presence of particle source

Ryoko Tatsumi, Kazuo Hoshino, Akiyoshi Hatayama

Computer Physics Communications （Elsevier {BV}）  264   107960 2021年07月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00104655

　概要を見る

Eulerian–Lagrangian (EL–LG) scheme is a numerical scheme that tracks pseudo particles in Eulerian cells. It is widely used in the computational fluid dynamics, however, numerical errors associated with a particle source term has not yet been investigated much. Hence this study focuses on numerical errors of EL–LG caused by particle sources. The purposes are: (i) to clarify causes and situations that bring larger numerical errors by source terms and (ii) to suggest an idea to reduce them. For those purposes, we focus on the particle continuity equation and carry out systematic analysis of the numerical error by setting the following three simple cases: Case (A) No source, Case (B) Constant source, and Case (C) source with arbitrary spatial profile. For each case, we have obtained a theoretical expression of the numerical error. It has been clarified that the errors become relatively large when (i) the spatial profile of the particle source has a large gradient and (ii) the source is localized in the region with high flow-velocity. These were caused by the treatment of the particle source: If pseudo particles due to the source are added in a simple way at the start or the end of the time step, this can lead to larger numerical errors. To reduce those errors, a time-averaging scheme has been suggested. Although the analyzed cases are simple, the results obtained in this study would give important knowledge and insight into numerical errors associated with particle sources in EL–LG schemes.

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Numerical Simulation Study of the Magnetic Flux Tube Expansion on the Divertor Plasma Parameters by the LINDA Code

ISLAM Md. Shahinul, NAKASHIMA Yousuke, ISHIGURO Seiji, HOSHINO Kazuo, HATAYAMA Akiyoshi, HASEGAWA Hiroki, SAKAMOTO Mizuki

Plasma and Fusion Research （一般社団法人 プラズマ・核融合学会）  16 （ 0 ） 2403049 - 2403049 2021年03月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  18806821

　概要を見る

<p>In this research, we investigate the effect of magnetic flux tube expansion on the divertor plasma parameters by using the fluid code "LINDA". A comparison between the cylindrical flux tube (without the magnetic flux expansion) and the expansion magnetic flux tube has been undertaken. The aim of the study is to understand the impact of magnetic field expansion on the divertor physics by using the LINDA fluid code. The plasma density (<i>n</i>_i) is decreased and parallel velocity is increased (<i>u</i>_i||) toward the target plate with the expansion of magnetic field lines near the target plate. The heat and particle fluxes are reduced significantly on the target plate in the case of the expansion mesh configuration. For the case of cylindrical mesh, advection becomes stronger with the decreasing distance from the target plate. In the case of expansion mesh, diffusion is stronger with the decreasing distance from the target plate. These outcomes clearly indicate the effect of the magnetic field structure on the divertor plasma parameters.</p>

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Simulation studies of divertor detachment and critical power exhaust parameters for Japanese DEMO design

Asakura N., Hoshino K., Homma Y., Sakamoto Y.

Nuclear Materials and Energy 26   100864 2021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

　概要を見る

© 2020 The Author(s) Handling of a large thermal power exhausted from the confined plasma is one of the most important issues for ITER and DEMO. A conventional divertor, which has the closed geometry similar to that of ITER and longer leg of 1.6 m, was proposed for the Japanese (JA) DEMO reactor (Rp/ap = 8.5/2.42 m). A radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated by SONIC simulation, in order to evaluate the power exhaust in JA-DEMO 2014 (primary design with Psep ~ 283 MW) and JA-DEMO with higher plasma elongation (a revised design with Psep ~ 235 MW). The divertor operation with the peak qtarget ≤ 10 MWm−2 was determined in the low nesep of 2–3 × 1019 m−3 under the severe conditions of reducing radiation loss fraction, i.e. f*raddiv = (Pradsol + Praddiv)/Psep, and diffusion coefficients (χ and D). The divertor geometry and reference key parameters (f*raddiv ~ 0.8, χ = 1 m2/s and D = 0.3 m2/s) were so far consistent with the power exhaust concepts in the nesep range, and the revised JA-DEMO design has advantages of wider nesep range and enough margin for the divertor operation. For either severe assumption of f*raddiv ~ 0.7 or χ and D to the half value, higher nesep operation was required for the primary design in order to control the peak qtarget ≤ 10 MWm−2, i.e. the operation window was reduced. Applying the two severe assumptions, the divertor operation was difficult in the low nesep range for the both designs.

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Detached helium plasma simulation by a one-dimensional fluid code with detailed collisional-radiative model

Tanaka H., Saeki I., Ohno N., Kajita S., Ido T., Natsume H., Hatayama A., Hoshino K., Sawada K., Goto M.

Physics of Plasmas 27 （ 10 ） 102505 2020年10月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  1070664X

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© 2020 Author(s). To increase the accuracy of a particle, momentum, and energy source terms in the detached helium plasma simulation, rate coefficients with the collisional-radiative model were introduced into the fluid code LINear Divertor Analysis (LINDA). Obtained effective rate coefficients and related source terms were compared with those from the conventional empirical databases. It is shown that a high-density condition in future fusion devices causes larger deviation between the effective and the empirical source terms. One-dimensional detached plasma simulation indicated that the peak amplitude of the plasma density during the rollover is sensitive to the source term difference related to the recombination. This study additionally revealed that the heating effect in the three-body recombination process strongly affects the detached plasma formation and downstream plasma parameters.

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Study of the Transient Behavior of Detached Plasma during Xe Gas Injection into the D-Module of GAMMA 10/PDX

ISLAM Md. Shahinul, HOSHINO Kazuo, HATAYAMA Akiyoshi, HASEGAWA Hiroki, ISHIGURO Seiji, MATSUURA Hiroto, SAKAMOTO Mizuki, NAKASHIMA Yousuke, IIJIMA Takaaki, NOJIRI Kunpei, EZUMI Naomichi, YOSHIKAWA Masayuki, KARIYA Tsuyoshi, MINAMI Ryutaro, HIRATA Mafumi

Plasma and Fusion Research （一般社団法人 プラズマ・核融合学会）  15 （ 0 ） 1402074 - 1402074 2020年10月

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

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<p>The transition phenomenon of detached plasma during additional heating effects has been studied by applying a short pulse (25 ms) of Electron Cyclotron Heating (ECH) at the east plug-cell of the tandem mirror device GAMMA 10/PDX. In this paper, the plasma parameters on the target plate of the D-module are studied for understanding the impact of transient heating pulse on the sustainability of the detached plasma. The heat flux increases on the target plate for the additional heating case in comparison to without heating pulse. The ion flux also increases when the ECH heating is activated. The electron density enhances significantly during the ECH heating pulse application period. Furthermore, the electron density and the ion flux increase according to the increment of Xe plenum pressure in the case of ECH heating condition. The experimental results clarified that ECH heating pulse can drive the detachment state to attached state in spite of strong effect of Xe gas for generating the detached plasma.</p>

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Study of multiple impurity seeding effect using SONIC integrated divertor code for JT-60SA plasma prediction

Yamoto S., Hoshino K., Homma Y., Nakano T., Hayashi N.

Contributions to Plasma Physics 60 （ 5-6 ） e201900146 2020年07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  08631042

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© 2019 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim In order to study the potential impurity seeding operation regime of the future fusion devices, the first application of the integrated divertor code SONIC to the Ar + Ne mixed-impurity seeding operation of JT-60SA steady-state high-β plasma has been carried out. In the case, Ne is added to Ar-only seeding, the separatrix electron density has fell into the desired low separatrix electron density of the scenario. This is mainly because the D+ flow velocity towards the inner divertor has been increased by the Ne seeding. The resultant friction force transports Ar impurities towards the inner divertor region, while impurities are stagnated in the top of scrape-off layer (SOL) in the Ar-only seeding case. The higher impurity radiation power in the divertor regions and lower one in the SOL region above the X point have been obtained in mixed-impurity seeding cases, which show similar tendency as the Ar + Ne mixed-impurity seeding experiment in JT-60 U. At the core edge, Zeff has been slightly increased and the radiation power has been decreased as the Ne seeding rate increases. The core plasma/impurity transport has been also evaluated by the TOPICS code using the impurity density at the core edge computed by the SONIC as a boundary parameter. The results show lower Zeff and radiation power, and higher electron temperature in the core in the mixed-impurity seeding cases. Above possible contributors to the better energy confinement indicate that the mixed-impurity seeding operation might be more effective than Ar-only seeding operation.

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First EMC3-EIRENE modelling of JT-60SA edge plasmas with/without resonant magnetic perturbation field

Tanaka H., Kawamura G., Hoshino K., Kobayashi M., Matsunaga G., Suzuki Y., Lunt T., Feng Y., Ohno N.

Contributions to Plasma Physics 60 （ 5-6 ） e201900114 2020年07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  08631042

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© 2020 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim The first EMC3-EIRENE modelling for the JT-60SA edge plasmas with/without the resonant magnetic perturbation field (RMP) was conducted. Both core degradation in the vacuum approximation and strike-point splitting were observed due to the application of the RMP, as in previous studies in other devices. Toroidally averaged divertor heat fluxes with and without the RMP were fitted by using a one-dimensional diffusive model function, showing similar profiles. To extract the modulation components of the divertor heat flux profile, the proper orthogonal decomposition (POD) method was applied. By using the POD outputs, a modulation ratio with respect to the toroidally averaged profile was quantitatively calculated.

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Comparison of the Process Systems Code with the SONIC Divertor Code

Morris J., Asakura N., Homma Y., Hoshino K., Kovari M.

IEEE Transactions on Plasma Science 48 （ 6 ） 1799 - 1803 2020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  00933813

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© 1973-2012 IEEE. In a demonstration (DEMO) reactor, mitigation of the large heat load on the divertor target to the below material and engineering limits is a key requirement for operation. Systems modeling is used to design entire fusion power plants and, therefore, has to be able to appropriately capture the divertor challenge. Therefore, it is important to validate these models against comprehensive SOL-divertor simulation codes and experiments. A 1-D divertor model in PROCESS was investigated, compared to the results of 2-D SONIC simulation under the detachment condition. The comparison shows how the 1-D divertor model handles the power loss mechanisms from the outboard mid-plane to the outer divertor target for a DEMO-like condition. The results show good agreement on the calculated value of the total power crossing the separatrix (<5% difference) and the total impurity radiation power P imp (<10% difference). However, the 1-D profiles show differences in density and temperature at the upstream of the target (<10 m of connection length to target, corresponding to 10 cm in the poloidal length). One reason for this difference is that the 2-D model calculates impurity transport, which produces a variable impurity fraction along the connection length in the divertor, while the 1-D model uses a single averaged value. The SONIC code also considers physical processes not covered in the 1-D model, such as radial transport in the SOL and divertor region. A scan of P sep values in PROCESS for a DEMO-sized machine show that above Psep = 200 MW, there is a stronger impact on cost and machine size for higher Psep.

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Effect of collisionality dependence of thermal force on impurity transport under a lower collisional condition in DEMO scrape-off layer plasma

Homma Y., Hoshino K., Yamoto S., Tokunaga S., Asakura N., Sakamoto Y.

Nuclear Fusion 60 （ 4 ） 046031 2020年03月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00295515

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© 2020 IAEA, Vienna. A preceding study (Homma et al 2018 Contrib. Plasma Phys. 58 629-37) presented an advanced model for thermal force whose applicable condition is extended from collisional to relatively low collisional plasma. According to this extended model, the thermal force reduces when plasma collisionality decreases. In the present study, the extended thermal force model has been implemented into a SOL-divertor integrated simulation code SONIC, in order to study the impact on impurity transport in DEMO-relevant scrape-off layer (SOL) plasma, due to thermal force reduction. A set of test simulations has been carried out supposing a reference steady-state operation scenario of the Japanese DEMO fusion reactor concept (JA DEMO). The thermal force has reduced by as much as 20%-70%, reflecting relatively lower collisionality in DEMO SOL plasma. The simulation results have demonstrated that the introduction of the collisionality dependence of thermal force leads to as much as 20%-80% of effective decrease in impurity density and its content ratio widely over the SOL upstream area under the DEMO relevant condition. Compared to the case with the conventional thermal force model, relative change rate of impurity content at representative poloidal positions are as follows: low-field side (LFS) X-point-59%, poloidal top area-22%, HFS upstream area up to-80%, and no major impact around LFS upstream.

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IMPGYRO: The full-orbit impurity transport code for SOL/divertor and its successful application to tungsten impurities

Yamoto S., Homma Y., Hoshino K., Toma M., Hatayama A.

Computer Physics Communications 248   106979 2020年03月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00104655

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© 2019 Elsevier B.V. In future fusion reactors, tungsten (W) is currently regarded as one of the most feasible candidates for the plasma-facing components (PFCs). Once the W impurities sputtered from the PFCs penetrate into the core plasma, the large radiation cooling and fuel dilution, which leads to the deterioration of the core plasma performance, take place. To understand and to control W impurity transport are indispensable for future fusion reactors. In order to understand W transport, we are continuing to develop the kinetic impurity transport code IMPGYRO for SOL/divertor in fusion reactors. The IMPGYRO has obtained following unique features compared to other existing kinetic impurity transport codes through the 15-years development activity; (i) the exact Larmor motion of impurity ions are computed so that the effects of drifts are automatically taken into account, (ii) the Coulomb collision between impurities and background plasma ions are modelled by the Binary Collision Method which kinetically calculates more precise friction and thermal forces, and (iii) the background plasma transport and impurity transport are computed self-consistently by the coupled calculation with SOL/divertor plasma transport code such as SOLPS-ITER. Although the IMPGYRO transport model has significantly improved from the initial development phase, the modelling improvement of the IMPGYRO has never been described in detail. In this paper, (a) the current W generation/transport model, and (b) the coupling strategy between the SOLPS-ITER and the IMPGYRO are summarized. The new results (c) the check of the prompt re-deposition model of the IMPGYRO, and (d) the further extension of the Coulomb collision modelling of the IMPGYRO towards the neoclassical (NC) transport handling capability has been performed. The effects of the NC transport processes on a simple circular tokamak geometry are compared with the anomalous diffusion by means of the extended IMPGYRO to obtain better understanding of cross-field transport process of W. The results show that the NC transport process may become non-negligible compared to the anomalous diffusion at least in this configuration. Based on these recent improvements, the present version of the IMPGYRO is very useful tool for understanding and prediction of W transport in SOL/divertor regions of fusion devices.

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Simulation study of mixed-impurity seeding with extension of integrated divertor code SONIC

Yamoto S., Hoshino K., Homma Y., Nakano T., Hayashi N.

Plasma Physics and Controlled Fusion 62 （ 4 ） 045006 2020年02月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  07413335

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© 2020 IOP Publishing Ltd. Aimed at obtaining key physics that determine the controllability of impurity transport in the scrape-off layer (SOL)/divertor regions, the integrated divertor code SONIC has been further extended to handle three or more impurity species kinetically. The extended SONIC code has been applied to the steady-state high-beta scenario-like plasma of JT-60SA as a testbed. We first performed a Ne transport simulation on the fixed Ar-seeded background plasma. Different radiation power distribution along the magnetic field line was obtained between Ar and Ne. The Ar radiation is strong around the top region of the SOL, which is mainly due to the line radiation of highly charged Ar ions trapped by the thermal force. In contrast, the Ne radiation is strong around the high-field side near the X point, mainly due to the line radiation of Ne7+ trapped by the balance between the thermal force and the frictional force with D+ parallel flow. We performed a parametric survey of Ne seeding rate as a second step. The effects of Ne transport on the plasma are self-consistently computed. The Ne impurities are injected into the plasma with a fixed puff rate of Ar. Even a small Ne seeding rate of 0.02 Pa m3 s-1 results in lower Ar radiation power in the SOL and core edge than in the Ar-only case. This is mainly due to the high D+ parallel flow velocity towards the inner divertor in the Ar + Ne seeding case. The resultant frictional force transports the Ar impurities towards the inner divertor region. When the line radiation of Ne7+ is switched off in the simulation, such high D+ parallel flow cannot be seen. These results suggest that the line radiation of Ne7+ has a key role for the high D+ parallel flow. The results show the possibility of impurity transport control in the SOL by mixed-impurity seeding.

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Development of a Three Dimensional Plasma Fluid Code Using a Lagrange-Monte-Carlo Scheme

TATSUMI Ryoko, RUNOV Alexei, SCHNEIDER Ralf, HOSHINO Kazuo, HATAYAMA Akiyoshi

Plasma and Fusion Research （社団法人 プラズマ・核融合学会）  15 （ 0 ） 1403003 - 1403003 2020年02月

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

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<p>We are developing a Lagrange (LG)-Monte-Carlo (MC) scheme for three-dimensional (3D) SOL/Divertor plasma fluid modeling. By using test particles, the scheme is suitable for handling 3D complex geometries. The semi-implicit treatment of the pressure gradient term enables us to improve the robustness of the coupling of the continuity and the momentum equations. Detailed numerical checks of the integrated scheme of LG-MC have been done for a simple 1D geometry. Benchmark tests between the new LG-MC and a conventional Finite-Volume scheme were carried out and good agreement was obtained. A first test calculation for a 3D cylindrical geometry has been also successfully done.</p>

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Global modelling of tungsten impurity transport based on the drift-kinetic equation

Kanno R., Kawamura G., Nunami M., Homma Y., Hatayama A., Hoshino K.

Nuclear Fusion 60 （ 1 ） 016033 2019年12月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00295515

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© 2019 IAEA, Vienna. A global kinetic simulation model of collisional impurity transport is developed for evaluating the radial particle flux of tungsten impurity in the edge region of a tokamak plasma. Here, the plasma including the impurity and the background ion is presupposed to be quasi-steady. The simulation model is based on the drift-kinetic equation of the impurity affected by the friction force and the thermal force, which were formulated in the previous study (Homma et al 2016 Nucl. Fusion 56 036009). The model is implemented in a drift-kinetic δf simulation code. We find that the magnetic drift term in the drift-kinetic equation causes the 'global effect' on the impurity transport. Here, the global effect means that the solution δfz of the drift-kinetic equation (and also the radial particle flux) on a magnetic flux surface is influenced by the values of the solution δfz all over the edge region.

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Numerical Simulation for Enhancement of H⁻Production in the DC Arc-Discharge Hydrogen Negative Ion Source for Medical Use

YAMADA Shota, KITAMI Hisashi, NOMURA Shinji, AOKI Yasushi, HOSHINO Kazuo, HATAYAMA Akiyoshi

Plasma and Fusion Research （社団法人 プラズマ・核融合学会）  14 （ 0 ） 3401160 - 3401160 2019年10月

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

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<p>A multi-cusp DC arc-discharge hydrogen negative ion (H⁻) source has been developed for proton cyclotron, which is used for Boron Neutron Capture Therapy (BNCT). The purpose of this study is to understand the dependence of H⁻production on the operation parameters and optimize H⁻production in the source. In this paper, we focus on the effect of magnetic filter field on H⁻volume production. The H⁻density in the extraction region has been numerically estimated in the three patterns of magnetic filter field by using 3D electron transport code and 0D rate equations. The result suggests that the magnetic filter field which is localized in the vicinity of the extraction region is suitable for the efficient H⁻volume production.</p>

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Study of end-cell plasma parameters of GAMMA 10/PDX by the LINDA code

Islam M., Nakashima Y., Hatayama A., Ishiguro S., Hoshino K., Ezumi N., Sakamoto M.

Plasma Physics and Controlled Fusion 61 （ 2019 ） 125005 2019年10月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  07413335

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© 2019 IOP Publishing Ltd. This research numerically investigates the plasma behavior during neutral gases puffing into the west end-cell of the linear plasma confinement device GAMMA 10/PDX. A multi-fluid code named 'LINDA' is developed and applied at the west end-cell of the device to explore the processes of plasma detachment and energy loss mechanism during plasma-neutral interactions. The ion heat flux limiter is newly introduced in the LINDA code. In the paper, we investigate the influence of the ion heat flux limiter in our divertor plasma models to understand how the ion heat flux limiter affects the computed plasma parameters. The upstream plasma parameters are ni = 1 × 1019 m-3,Te = 30 eV and Ti = 100 eV. The effect of ion heat flux limiting reduces the ion temperature and heat flux in the direction of the target plate. The ion temperature profile is shown to be similar when the heat flux limiting factor (α i) is varied from 0.2 to 2.0. The effect of Hydrogen (H) and Argon (Ar) gas puffing has been investigated under the condition of ion heat flux limiting factors α i = 1.0. The ion temperature is shown to be similar between Ar and without Ar puffing. On the other hand, T e reduces remarkably for Ar puffing by enhancing the radiation cooling of Ar. The ion and electron loss are greatly enhanced for combined puffing of Ar and H. The ionization loss is found to be slightly higher and the charge-exchange loss is found to be slightly lower during the ion heat flux limiting case. These outcomes may contribute to understanding the plasma detachment processes in fusion devices.

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Advances in the physics studies for the JT-60SA tokamak exploitation and research plan

Giruzzi G., Yoshida M., Aiba N., Artaud J.F., Ayllon-Guerola J., Beeke O., Bierwage A., Bolzonella T., Bonotto M., Boulbe C., Chernyshova M., Coda S., Coelho R., Corona D., Cruz N., Davis S., Day C., De Tommasi G., Dibon M., Douai D., Farina D., Fassina A., Faugeras B., Figini L., Fukumoto M., Futatani S., Galazka K., Garcia J., Garcia-Muñoz M., Garzotti L., Giudicotti L., Hayashi N., Honda M., Hoshino K., Iantchenko A., Ide S., Inoue S., Isayama A., Joffrin E., Kamada Y., Kamiya K., Kashiwagi M., Kawashima H., Kobayashi T., Kojima A., Kurki-Suonio T., Lang P., Lauber P., De La Luna E., Marchiori G., Matsunaga G., Matsuyama A., Mattei M., Mazzi S., Mele A., Miyata Y., Moriyama S., Morales J., Moro A., Nakano T., Neu R., Nowak S., Orsitto F.P., Ostuni V., Oyama N., Paméla S., Pasqualotto R., Pégourié B., Perelli E., Pigatto L., Piron C., Pironti A., Platania P., Ploeckl B., Ricci D., Romanelli M., Rubino G., Sakurai S., Srkimki K., Scannapiego M., Shinohara K., Shiraishi J., Soare S., Sozzi C., Suzuki T., Suzuki Y., Szepesi T., Takechi M., Tanaka K., Tojo H., Turnyanskiy M., Urano H., Valisa M., Vallar M., Varje J., Vega J., Villone F., Wakatsuki T., Wauters T., Wischmeier M.

Plasma Physics and Controlled Fusion （Plasma Physics and Controlled Fusion）  62 （ 2020 ） 014009 2019年10月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  07413335

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© Gerardo Giruzzi. JT-60SA, the largest tokamak that will operate before ITER, has been designed and built jointly by Japan and Europe, and is due to start operation in 2020. Its main missions are to support ITER exploitation and to contribute to the demonstration fusion reactor machine and scenario design. Peculiar properties of JT-60SA are its capability to produce long-pulse, high-β, and highly shaped plasmas. The preparation of the JT-60SA Research Plan, plasma scenarios, and exploitation are producing physics results that are not only relevant to future JT-60SA experiments, but often constitute original contributions to plasma physics and fusion research. Results of this kind are presented in this paper, in particular in the areas of fast ion physics, high-beta plasma properties and control, and non-linear edge localised mode stability studies.

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Effects of the extraction voltage on the beam divergence for a H^- ion source

Lindqvist M., Nishioka S., Miyamoto K., Hoshino K., Lettry J., Hatayama A.

Journal of Applied Physics 126 （ 2019 ） 123303 2019年09月

研究論文（学術雑誌）, 共著, 査読有り,  ISSN  00218979

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© 2019 Author(s). Negative hydrogen (H-) ion sources have a wide range of applications. The general requirement for these H-ion sources is to produce intense H-ion beams with good beam optics. The purpose of this paper is to clarify the effects of the beam extraction voltage on the beam divergence angle by three-dimensional Particle-in-Cell (PIC) modeling. Perveance matching has been studied for a wide range of the extraction voltage with the model geometry of a Linac4 H-ion source. The extracted H-beam divergence angle is evaluated for extraction voltages ranging from 7 to 14 kV by using the Keio-BFX PIC code. The results show divergence minima in the range of 9.2-11.5 kV for the case without surface H-production, which correspond to experimental results. The dependence of divergence on the extraction voltage is explained by the change of the shape of the meniscus. In particular, a flat meniscus corresponds to low beam divergence, and particles extracted through the center of the meniscus have a lower divergence than particles extracted near the Plasma Electrode.

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Effects of the Plasma Blob Nonlinear Formation/Transport on Impurity Transport in the SOL Regions

MAEDA Tomoyuki, HASEGAWA Hiroki, ISHIGURO Seiji, HOSHINO Kazuo, HATAYAMA Akiyoshi

Plasma and Fusion Research （社団法人 プラズマ・核融合学会）  14 （ 0 ） 3403133 - 3403133 2019年07月

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

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<p>The purpose of this study is to make clear the effects of the plasma blob on the impurity transport in the Scrape-Off Layer (SOL) regions. By using the two-dimensional (2D) interchange turbulence model with plasma transport along magnetic field line, we have simulated the plasma blob formation and transport in the SOL regions. We have studied the effects of the plasma blob on the impurity transport by using impurity equations of motion in the trace impurity limit. The results show that the plasma blob transport direction is not only the radial direction but also poloidal direction due to the plasma transport along the magnetic field line. Also, the impurity ions are transported radially outward with the plasma blob transport and the velocity of the impurity ions transport is the same order to the plasma blob velocity radially outward. These results implied that the impurity transport with the plasma blob may possibly be not only diffusion but also convection. Moreover, this motion of the impurity ions seems like eddy motion. As a next step, we study relation between this eddy motion of impurity ions and the plasma blob motion. That is our future work.</p>

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Analysis of Molecular Activated Recombination in Detached Divertor Plasmas

TSUBOTANI Yuka, TATSUMI Ryoko, HOSHINO Kazuo, HATAYAMA Akiyoshi

Plasma and Fusion Research （社団法人 プラズマ・核融合学会）  14 （ 0 ） 2403108 - 2403108 2019年07月

研究論文（学術雑誌）, 共著, 査読有り

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<p>Molecular activated recombination (MAR) has been received attention as one of the most important processes to realize the detached state, specifically in the linear plasma device, GAMMA10/PDX. The impact of ELMs on the MAR reaction and the resultant detached state, however, have not fully been understood. We have developed a zero-dimensional model of atomic-molecular processes which takes into account MAR by separate treatment of vibrational excited states of H₂molecules, and the high energy tail of the electron energy distribution function (EEDF) during the ELM heat pulse. The results show that the rate of MAR increases during ELMs due to an increase in excited molecules.</p>

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Investigation of Dependence of Power Transfer on Frequency in the Linac4 Negative Hydrogen Ion Source

KOBAYASHI Wakaba, NISHIDA Kenjiro, NOLL Daniel, LETTRY Jacques, HOSHINO Kazuo, HATAYAMA Akiyoshi

Plasma and Fusion Research （社団法人 プラズマ・核融合学会）  14 （ 0 ） 3401094 - 3401094 2019年06月

研究論文（学術雑誌）, 共著, 査読有り

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<p>Linac4 is radio frequency (RF) driven type negative hydrogen H⁻ion source. In the RF system, it is required to match the impedance between the RF-system and plasma loading. In order to match the impedance, it is essential to analyze the plasma impedance. In this study, we analyze the dependence of power transfer on the input parameters (frequency, input power and H₂gas pressure) by using the plasma impedance prediction model. The results indicate that in the case of frequency = 2 MHz, which is often used in the experiment, efficient power supply is seen in the wide range of input power and H₂gas pressure. Also, if the optimum value of input power and H₂gas pressure is input, it may be possible to reduce the influence of the frequency value on the power reflection efficiency.</p>

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Analysis of Ar Impurity Transport in the Large Tandem Mirror Device GAMMA 10/PDX Plasmas

TAKECHI Sayuri, TATSUMI Ryoko, ISLAM Md. Shahinul, HOSHINO Kazuo, HATAYAMA Akiyoshi, NAKASHIMA Yousuke

Plasma and Fusion Research （社団法人 プラズマ・核融合学会）  14 （ 0 ） 2403045 - 2403045 2019年03月

研究論文（学術雑誌）, 共著, 査読有り

　概要を見る

© 2019 The Japan Society of Plasma Science and Nuclear Fusion Research. The purpose of this study is to understand the Ar impurity transport process in the GAMMA 10/PDX. In the present study, the initial simulation using LINDA code and IMPGYRO code has been done to establish the basis of further understanding of the Ar impurity transport. The validity of the simulation model has been discussed by comparisons with those by the theoretical estimations of the friction force and the thermal force. The simulation results show that the Ar impurities with Z = 1 are transported towards the core region, since the thermal force dominates over the friction force under the present calculation conditions. The tendency does not contradict the experimental results and is reasonably explained by the theoretical prediction, i.e., the force balance between the thermal force and the friction force. The robust basis for further model validation and understanding of the Ar transport in the GAMMA 10/PDX has been established.

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Analysis of plasma impedance in the Linac4 negative hydrogen ion source

Kobayashi W., Nishida K., Mattei S., Lettry J., Hoshino K., Hatayama A.

AIP Conference Proceedings 2052   050009 2018年12月

研究論文（国際会議プロシーディングス）, 共著, 査読有り

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© 2018 Author(s). Linac4 is radio frequency (RF) driven type negative hydrogen H- ion source. In the RF system, it is required to match the impedance between the RF-system and plasma loading. In order to match the impedance, it is essential to analyze the plasma impedance. In this study, we analyze the plasma impedance by using the Electro-Magnetic Particle in Cell model. As a result, plasma impedance calculated by the Electro-Magnetic Particle in Cell model agrees qualitatively with one calculated by fluid model model in the steady state. On the other hand, plasma impedance calculated by the numerical simulation code is significantly different from one calculated by the zero dimensional equivalent circuit model during plasma ignition. The results show that the kinetic effects are necessary especially during plasma ignition. It is useful to propose more efficiency power transfer.

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Numerical simulation for the development of DC arc-discharge hydrogen negative ion source for medical use

Yamada S., Aoki Y., Takahashi N., Kitami H., Yaguchi M., Hoshino K., Hatayama A.

AIP Conference Proceedings 2052   060004 2018年12月

研究論文（国際会議プロシーディングス）, 共著, 査読有り

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© 2018 Author(s). Multi cusp DC arc-discharge hydrogen negative ion (H-) source has been developed for the proton cyclotron applications. The purpose of this study is to understand the dependence of the H- production on the operation parameters and optimize the H- production in the source. In this paper, we focus on the effect of magnetic filter field on the H- volume production. The H- volume production in the source has been numerically calculated in the two case of magnetic filter field by using 3D electron transport code and 0D rate equations. The result has suggest that the effect of magnetic filter field on H- beam current could be explained by the enhanced electron density in the extraction region.

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Plasma exhaust and divertor studies in Japan and Europe broader approach, DEMO design activity

Asakura N., Hoshino K., Utoh H., Someya Y., Suzuki S., Bachmann C., Reimerdes H., Wenninger R., Kudo H., Tokunaga S., Homma Y., Sakamoto Y., Hiwatari R., Tobita K., You J., Federici G., Ezato K., Seki Y., Ueda Y., Ohno N.

Fusion Engineering and Design 136   1214 - 1220 2018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

　概要を見る

© 2018 Elsevier B.V. Power exhaust scenario and divertor design for a steady-state Japan (JA) DEMO and a pulse Europe (EU) DEMO1 have been investigated as one of the most important common issues in Broader Approach DEMO Design Activity. Radiative cooling is a common approach for the power exhaust scenario. For the JA DEMO, development of the divertor design appropriate for high P sep /R p ∼30 MW m −1 is required, while the radiation fraction in the main plasma (f radmain = P radmain /P heat ) is ITER-level (0.40–0.45) and the exhaust power above the L- to H-mode power threshold (f LH = P sep /P thLH ) is large margin (∼2). For the EU DEMO1, larger f radmain (=0.67) and smaller f LH (=1.2) plasma is required, using higher-Z impurity seeding, in order to apply ITER-level divertor (P sep /R p = 17 MW m −1 ). ITER technology, i.e. water cooling with W-monoblock and Cu-alloy (CuCrZr) heat sink, is a baseline for JA and EU to handle the peak heat load of 10 MW m −2 -level, and neutron flux and irradiation dose are comparable. For the JA DEMO, two different water-cooling pipes, i.e. CuCrZr and F82H steel, are proposed. For the EU DEMO1, the heat sink consists of all Cu-alloy pipe, and the divertor size is reduced with replacing the baffles by the breeding blankets. Choices of the heat sink components have been developed appropriate to the high irradiation dose condition. These JA and EU approaches of the power exhaust scenario will provide important case studies for the future decision of the DEMO divertor design.

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Overview of the DEMO conceptual design activity in Japan

Tobita K., Hiwatari R., Utoh H., Miyoshi Y., Asakura N., Sakamoto Y., Someya Y., Homma Y., Nakamura M., Hoshino K., Tanigawa H., Nakamichi M., Tokunaga S., Kudo H., Nishimura A.

Fusion Engineering and Design 136   1024 - 1031 2018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

　概要を見る

© 2018 Elsevier B.V. The requirements for Japanese DEMO are to demonstrate (1) steady and stable electrical power generation, (2) plant availability on a level relevant for commercialization, and (3) self-sufficiency of fuels. The DEMO design activity has been mainly focused on challenging DEMO design issues such as divertor heat removal, superconducting magnets, breeding blankets, remote maintenance and the water cooling system, in order to reach the systemic consistency toward a promising DEMO concept. The current status of the DEMO conceptual design activity is presented.

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Predictive modelling of JT-60SA high-beta steady-state plasma with impurity accumulation

Hayashi N., Hoshino K., Honda M., Ide S.

Nuclear Fusion 58 （ 6 ） 066001 2018年06月

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

　概要を見る

© 2018 IAEA, Vienna. The integrated modelling code TOPICS has been extended to include core impurity transport, and applied to predictive modelling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce the divertor heat load. In the modelling, models and conditions are selected for a conservative prediction, which considers a lower bound of plasma performance with the maximum accumulation of impurity. The conservative prediction shows the compatibility of impurity seeding with core plasma with high-beta (β N > 3.5) and full current drive conditions, i.e. when Ar seeding reduces the divertor heat load below 10 MW m-2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate for Ar radiation. Due to the strong dependence of accumulation on the pedestal density gradient, high separatrix density is important for the low accumulation as well as the low divertor heat load. The conservative prediction also shows that JT-60SA has enough capability to explore the divertor heat load control by impurity seeding in high-beta steady-state plasmas.

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Study of mirror effect on scrape-off layer-divertor plasma based on a generalized fluid model incorporating ion temperature anisotropy

Togo S., Takizuka T., Reiser D., Hoshino K., Ibano K., Li Y., Ogawa Y., Sakamoto M., Ezumi N., Nakashima Y.

Contributions to Plasma Physics 58 （ 6-8 ） 556 - 562 2018年05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  08631042

　概要を見る

© 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim The mirror effect on scrape-off layer (SOL)-divertor plasma profiles was studied by relaxing the assumption of homogeneous magnetic fields used in a one-dimensional (1D) plasma fluid model, which incorporates the anisotropic ion temperature. For collisionless plasmas, the perpendicular ion temperature proportional to the magnetic field strength is observed, which reflects the conservation of the magnetic moment. In diverging field cases, the plasma flow becomes supersonic, while it is kept subsonic in contracting field cases. Parallel momentum gains/losses and energy transfers between parallel and perpendicular components caused by the mirror effect are also observed. In addition, it is indicated that the viscous flux approximation is invalid in inhomogeneous magnetic fields regardless of the plasma collisionality.

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An extended kinetic model for the thermal force on impurity particles in weakly collisional plasmas

Homma Y., Hoshino K., Tokunaga S., Yamoto S., Hatayama A., Asakura N., Sakamoto Y., Tobita K.

Contributions to Plasma Physics 58 （ 6-8 ） 629 - 637 2018年04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  08631042

　概要を見る

© 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim We have developed an extended kinetic model for the thermal force (Formula presented.) on test impurity particles in plasmas. This model makes use of a combination of the plasma ion heat flux q GM derived from the generalized moment equations (E. Zawaideh et al., Phys. Fluids 1988, 31) and the numerical algorithm presented earlier (Y. Homma, A. Hatayama, J. Comput. Phys. 2013, 250). The applicable condition extends from collisional to weakly collisional plasmas, and (Formula presented.) weakens when the plasma collisionality decreases. A test simulation is performed assuming a tentative weakly collisional plasma. The results show that it can have a non-negligible influence on transient impurity transport in a DEMO-scale tokamak to take into account the effect of plasma collisionality in thermal force modelling.

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Multi-impurity divertor simulation using a Monte Carlo kinetic impurity transport model

Hoshino K., Homma Y., Tokunaga S., Shimizu K.

Contributions to Plasma Physics 58 （ 6-8 ） 638 - 642 2018年04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  08631042

　概要を見る

© 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim The effects of multi-impurity transport are important for the divertor plasma performance. To simulate the radiative divertor plasma precisely, we have extended the integrated divertor code, SONIC, to the simulation with multi-impurity species. For SONIC extension, a new integrated modelling framework using the multiple-program multiple-data approach and message passing interface data exchange has been developed. In the framework, two impurities and more can be treated simultaneously with several Monte Carlo kinetic impurity codes. The extended SONIC code has been applied to analyse the JT-60SA radiative divertor plasma. The SONIC simulation including multi-impurity transport demonstrates that the radiative divertor plasma scenario compatible with the steady-state high-β plasma can be achieved.

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New model of plasma heat load on the first wall

Miyoshi Yuya, Asakura Nobuyuki, Hoshino Kazuo, Hiwatari Ryoji, Someya Youji, Takase Haruhiko, Sakamoto Yoshiteru, Tobita Kenji

Fusion Engineering and Design 124   267 - 270 2017年11月

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

　概要を見る

<p>Scrape Off Layer (SOL) plasma circumnavigates the torus along magnetic field lines, and some part of it encounters a leading edge of the first wall (FW). This leads the concentration of the plasma energy in a small area, causing a hot spot. In a DEMO reactor that deals with a few hundreds of MW of plasma energy in the SOL, the leading edge problem is a critical issue in the FW design. For the design purposes in DEMO and power reactors, a new analysis method is introduced, where a flux tube is poloidally divided into the Apple Peel Like Elements (APPLEs) and each APPLE is enclosed by adjacent 4 magnetic field lines. Considering the contact area of the APPLEs on the FW and the radial energy transport between the APPLEs, FW heat load profile can be analyzed. The result of the case that the fusion power is 1.5 GW, and the major radius is 8.2 m, shows that the heat load is peaked to about 1 MW/m² near the inboard midplane and the baffle plate. This method is expected to be useful for the DEMO design.</p>

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Physics design study of the divertor power handling in 8 m class DEMO reactor

Hoshino Kazuo, Asakura Nobuyuki, Tokunaga Shinsuke, Homma Yuki, Shimizu Katsuhiro, Sakamoto Yoshiteru, Tobita Kenji

Fusion Engineering and Design 124   352 - 355 2017年11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  0920-3796

　概要を見る

<p>The divertor plasma performance and the power handling are studied for 8 m class DEMO reactor with the fusion power of 1.5 GW. Due to the high impurity radiation power (80% of the exhausted power), the full detachment at the inner target and the partial detachment at the outer target are obtained for a relatively low electron density of 1.8 × 10¹⁹ m⁻³ at the outer mid-plane separatrix. The SONIC simulation shows the target heat load less than 8 MW/m², which can be handled by the ITER-like divertor target, for both target. However, at the outer target, the ion temperature is still high which may cause the target erosion. For the divertor power handling and suppression of the target erosion, the divertor design study have to be further proceeded as well as the core plasma design. Dependence of the mid-plane separatrix density and the impurity concentration on the fuel gas puff is also studied. With increasing fuel gas puff rate, the mid-plane separatrix density increases and the Ar impurity concentration at the outer mid-plane decreases to 0.5%.</p>

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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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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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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 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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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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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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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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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

　概要を見る

<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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Integrated modeling of whole Tokamak plasma

Hayashi Nobuhiko, Honda Mitsuru, Hoshino Kazuo, Hamamatsu Kiyotaka, Shimizu Katsuhiro, Takizuka Tomonori, Ozeki Takahisa, Fukuyama Atsushi

Plasma and Fusion Research 6   2403065 2011年12月

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

　概要を見る

<p>Development and integration of models for the whole tokamak plasma have progressed on the basis of experimental analyses and first principle simulations. Integrated models of core, edge-pedestal and scrape-offlayer (SOL)-divertor clarified complex and autonomous features of reactor relevant plasmas. The integrated core plasma model including an anomalous transport of alpha particles by Alfven eigenmodes is developed in the core transport code TOPICS-IB and indicates the degradation of fusion performance. The integrated rotation model is developed in the advanced transport code TASK/TX and clarifies the mechanism of alpha particle-driven toroidal flow. A transport model of high-Z impurities is developed and predicts large inward pinch in a plasma rotating in the direction counter to the plasma current. TOPICS-IB is extended to include the edge-pedestal model by integrating with the stability code, simple SOL-divertor and pellet models, and clarifies the mechanism of pellet triggered ELM. The integrated SOL-divertor code SONIC is further integrated with TOPICS-IB and enables to study and design operation scenarios compatible with both the high confinement in the core and the low heat load on divertor plates.</p>

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Evaluation of heat and particle controllability on the JT-60SA divertor

Kawashima, H., Hoshino, K., Shimizu, K., Takizuka, T., Ide, S., Sakurai, S., Asakura, N.

JOURNAL OF NUCLEAR MATERIALS 415 （ 1 ） S948 - S951 2011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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Inward pinch of high-Z impurity in a rotating tokamak plasma: effects of atomic processes, radial electric field and Coulomb collisions

Hoshino, K., Toma, M., Shimizu, K., Nakano, T., Hatayama, A., Takizuka, T.

NUCLEAR FUSION 51 （ 8 ） 211 - 215 2011年08月

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

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Improvement of the detachment modelling in the SONIC simulation

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

JOURNAL OF NUCLEAR MATERIALS 415 （ 1 ） S549 - S552 2011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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Modelling of JT-60U Detached Divertor Plasma using SONIC code

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

Journal of Plasma and Fusion Research SERIES 9   592 - 597 2010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

First Steps Towards the Coupling of the IMPGYRO and SOLPS Codes to Analyze Tokamak Plasmas with Tungsten Impurities

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

CONTRIBUTIONS TO PLASMA PHYSICS 50 （ 3-5 ） 392 - 397 2010年05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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Accumulation Process of High-Z Impurity in Toroidal Rotating Tokamak Plasma

Hoshino, K., Takizuka, T., Nakano, T.

CONTRIBUTIONS TO PLASMA PHYSICS 50 （ 3-5 ） 386 - 391 2010年05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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Modelling of Ion Kinetic Effects for SOL Flow Formation

Takizuka, T., Hoshino, K., Shimizu, K., Yagi, M.

CONTRIBUTIONS TO PLASMA PHYSICS 50 （ 3-5 ） 267 - 272 2010年05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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Integrated Modeling for Control of Advanced Tokamak Plasma

T Ozeki, N Hayashi, M Honda, N Aiba, K Hamamatsu, K Shimizu, K Hoshino, T Takizuka, S Tokuda

Journal of Plasma and Fusion Research SERIES 8   1138 - 1142 2009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

Analysis of the high-Z impurity transport in a tokamak by the IMPGYRO code

M. Toma, K. Hoshino, K. Inai, M. Ishida, A. Hatayama,　K. Ohya

Journal of Plasma and Fusion Research SERIES 8   1116 - 1119 2009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

Numerical analysis of incident angle of heavy metal impurity to plasma facing components by IMPGYRO

K. Hoshino, M. Toma, M. Furubayashi, A. Hatayama, K. Inai, K. Ohya

JOURNAL OF NUCLEAR MATERIALS 390-391   168 - 171 2009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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Comparison of kinetic and fluid neutral models for attached and detached state

Furubayashi, M., Hoshino, K., Toma, M., Hatayama, A., Coster, D., Schneider, R., Bonnin, X., Kawashima, H., Asakura, N., Suzuki, Y.

JOURNAL OF NUCLEAR MATERIALS 390-91   295 - 298 2009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  0022-3115

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Coupled IMPGYRO-EDDY simulation of tungsten impurity transport in tokamak geometry

Toma, M., Hoshino, K., Inai, K., Furubayashi, M., Hatayama, A., Ohya, K.

JOURNAL OF NUCLEAR MATERIALS 390-91   207 - 210 2009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  0022-3115

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Kinetic modelling of impurity transport in detached plasma for integrated divertor simulation with SONIC (SOLDOR/NEUT2D/IMPMC/EDDY)

Shimizu, K., Takizuka, T., Ohya, K., Inai, K., Nakano, T., Takayama, A., Kawashima, H., Hoshino, K.

NUCLEAR FUSION 49 （ 6 ） 550 - 554 2009年06月

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

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Benchmarking kinetic and fluid neutral models with drift effectsle

Hoshino, K., Toma, M., Hatayama, A., Coster, D. P., Bonnin, X., Schneider, R., Kawashima, H., Asakura, N., Suzuki, Y.

CONTRIBUTIONS TO PLASMA PHYSICS 48 （ 1-3 ） 136 - 140 2008年03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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High-Z impurity transport code by Monte Carlo method in a realistic tokamak geometry - IMPGYRO

Hoshino, K., Noritake, M., Toma, M., Hatayama, A.

CONTRIBUTIONS TO PLASMA PHYSICS 48 （ 1-3 ） 280 - 284 2008年03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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Modeling of multi-dimensional impurity transport in a realistic tokamak geometry

Fukano, A., Noritake, A., Hoshino, K., Yamazaki, R., Hatayama, A.

JOURNAL OF NUCLEAR MATERIALS 363   211 - 215 2007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  0022-3115

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Numerical analysis of the SOL/divertor plasma flow with the effect of drifts

Hoshino, K., Hatayama, A., Asakura, N., Kawashima, H., Schneider, R., Coster, D.

JOURNAL OF NUCLEAR MATERIALS 363   539 - 543 2007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  0022-3115

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Effect of drifts on the plasma flow in the detachment

Hoshino, K., Hatayama, A., Kawashima, H., Asakura, N., Schneider, R., Coster, D.

CONTRIBUTIONS TO PLASMA PHYSICS 46 （ 7-9 ） 591 - 596 2006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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Effect of drifts on the high Mach flow associated with the divertor detachment

Hoshino, K, Hatayama, A, Schneider, R, Coster, D

JOURNAL OF NUCLEAR MATERIALS 337 （ 1-3 ） 276 - 280 2005年03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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Multi-dimensional impurity transport code by Monte Carlo method including gyro-orbit effects

Hyodo, I, Hirano, M, Miyamoto, K, Hoshino, K, Hatayama, A

JOURNAL OF NUCLEAR MATERIALS 313   1183 - 1187 2003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,  ISSN  0022-3115

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