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The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokka...

原型炉設計合同特別チームのこれまでの成果である日本の原型炉概念(JA DEMO)の基本設計として、核融合科学技術委員会が示した原型炉の目標（① 数十万kWを超える定常かつ安定した電力、② 実用に供しうる稼働率,③ 燃料の自己充足性を満足する総合的なトリチウム増殖）を満足する定常炉概念と今後のR&D課題について報告した。特に主な成果として、発電プラント設備を含む原型炉全体像の構築とそれ基づく正味電気出力の明確化、耐圧性とトリチウム生産性を両立するトリチウム増殖ブランケットの概念設計、及び原型炉の安全性確保に向けた分析した想定起因事象とその緩和方策について報告した。, 第28回 国際原子力機関（IAEA）核融合エネルギー会議（IAEA-FEC2020）

The conceptual design of Japan’s fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in-vessel components to the primary cooling system is evaluated to be 5.7 g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94 kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 11.77TBq/year/loop (318 Ci/year/loop), which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, a confinement concept of tritium release from PHTS is discussed under the condition of ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA.

The Remote Experimentation Centre (REC) in Japan has been preparing to replicate the full dataset of ITER over 10 000 km distance. In such a multi-site data repository environment, the data location informing service will be essential to find and retrieve the data efficiently. Considering the long latency time and the self sustainability of remote sites, the data location database should be running at each repository site. Multi-master asynchronous replication between cooperating databases will be essential to realize the remote experimental collaborations in fusion research. This study has investigated the functional differences of some relational databases and found that Postgres BDR has the expected database replication capabilities. Bi-directional replication (BDR) tests by using the LHD database and SNET revealed that the throughputs are sufficient for remote collaborations in fusion experiments.

Technical verification has been progressing for high efficiency data replication between ITER and the Remote Experimentation Centre (REC) in Japan. Transferring a huge amount of data simultaneously to multiple destinations might cause excessive loads and network bandwidth on the sender so that daisy-chained relay transfer would be a considerable solution. This study demonstrates how efficiently the replication relay could be realized for the next-generation fusion experiments, such as ITER and JT-60SA. All the LHD data were consecutively sent to the REC through the global loop path (GLP: Toki - Gifu - Tokyo - Amsterdam - New York - Los Angeles - Tokyo - Aomori - Rokkasho) on SINET5 L2VPN, whose round-trip time is almost 400 ms. MMCFTP was used for the data transferring application. In both the Japan domestic path and the GLP cases, every transfer shows a very stable flattop speed as the preset 8 Gbps. However, longer gap times were needed in MMCFTP initial negotiation to establish numerous sessions. The performance optimized NVMe and iSCSI striped storages have shown higher throughputs than the ITER estimated initial data rate of 2 GB/s. Those knowledge enable the design optimization of not only the sender/receiver servers with their storages but also the intermediate relay server system.

The Broader Approach (BA) activities consists of three major projects: the International Fusion Energy Research Center (IFERC) project, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, and the Satellite Tokamak Programme (STP, JT-60SA). These projects have been carried out to obtain basic data for the design of DEMO fusion reactor from 2007. For 8-year activities, the above projects could produce a set of fruitful results for the DEMO reactor. DEMO design activity has been conducted to build a set of DEMO design bases in accordance with a series of discussion between EU and JA. In the DEMO R&D activities, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. Some additional R&D subjects recommended by peer review comments have also been studied successfully in recent years. Regarding the IFMIF/EVEDA project, some main components of the accelerator facility been designed and tested. The validation test using EVEDA Lithium Test Loop (ELTL) was also completed successfully in October 2014.

Developing effective and fast data transfer system for the huge amount data between Europe and Japan is a critical issue for the ITER Remote Experimentation Center (REC). To implement the system, effective data transfer methods and wide bandwidth international network are required.This paper describes results of data transfer experiments. We have evaluated two data transfer methods: Packet Pacing and MMCFTP. By using Packet Pacing and 2.4 Gbps line, we achieved 2.2 Gbps data transfer from NIFS to IFERC. By using MMCFTP and 10 Gbps line, we achieved 2.5 Gbps data transfer from NIFS to Dublin, Ireland. Furthermore, by using MMCFTP and 100Gbps line, we successfully achieved the stable transmission of 1PB of data at approximately 84 Gbps, one of the world’s fastest transmission speeds.This paper also describes the upgrade plan of SINET (a Japanese academic backbone network), which is used for ITER and REC communications. SINET will be upgraded to the network based on 100-Gigabit Ethernet technology in April 2016. Furthermore, direct lines of 20 Gbps (10 Gbps × 2) between Japan and Europe will be introduced. These direct lines will reduce latency between Europe and Japan and will realize higher speed data transfer., •This paper presents fast data transfer experiments using packet pacing and mmcftp and results.•An upgrade plan of Japanese Academic Network SINET is also described.•To send the huge amount of data from ITER to the ITER Remote Experiment Center (REC), effective transfer method and super high-speed internationalnetwork are required.•This paper presents a progress of the investigation for fast transfer method.

Goals of Japan’s fusion demonstration (DEMO) reactor are to demonstrate (1) steady and stable electric power generation in a power plant scale, (2) self-sufficient production of fuel (tritium), and (3) reasonable availability using a remote maintenance scheme anticipated in a commercial plant. Main design parameters of JA DEMO are a plasma major radius of 8.5 m, fusion output of 1.5-2 GW, magnetic field on the plasma axis of 5.94 T. The superconducting coil system of the reactor consists of a central solenoid (CS), 7 poloidal field (PF) coils and 16 toroidal field (TF) coils. Regarding CS and PF coils, superconducting coil technology on DEMO is basically the same as that on the world largest fusion experimental reactor called ITER. In contrast, TF coils have a technology gap on magnetic energy and the resulting stress between ITER and DEMO due to their size and magnetic field. In particular, the necessity of higher design stress is critical for TF coils, requiring the development of high strength cryogenic steels surpassing the existing ones. The fundamental design strategy to mitigate tolerances in TF coil fabrication is also presented.

Magnetic flux coordinates are constructed from an analytic solution (Ito and Nakajima 2009 Plasma Phys. Control. Fusion 51 035007) for the reduced magnetohydrodynamics (MHD) equilibrium equations for high-beta tokamaks in the presence of poloidal and toroidal flows comparable to the poloidal sound velocity. The analytic solution indicates non-circular magnetic flux surfaces and transition between sub- and super-sonic poloidal flows. The magnetic flux coordinates for such non-circular magnetic flux surfaces are obtained for stability analysis. The flux coordinates are numerically obtained from the high order polynomial equations for the relation with the geometrical coordinates. As applications, pressure profiles in the poloidal direction on each flux surface, which become non-constant due to flow, and the flux average of the pressure are obtained. The transitions of the pressure profiles and the flux average of the pressure between sub- and super-sonic poloidal flows are discussed in relation with the radial force balance. The flux coordinates are also obtained analytically by expanding the coordinate relations with respect to the inverse aspect ratio.

In Rokkasho Japan, the International Fusion Energy Research Center (IFERC) project and the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project are on going under the Broader Approach framework. The IFERC project consists of three sub-projects; a fusion demonstration reactor (DEMO) Design and R&D Coordination Center, a Computational Simulation Center (CSC), and an ITER Remote Experimentation Center (REC). DEMO Design activity has been conducted by the IFERC project team in Rokkasho and home teams in EU and JA. In the DEMO R&D activity, five R&D tasks mainly of the blanket materials are carried out intensively. A supercomputer with 1.23 Pflops of LINPAC performance has been installed in December 2011, the operation started in January 2012. Discussion of overall plan of REC has started in 2012 between EU and Japan. In the IFMIF/EVEDA project, an IFMIF prototype accelerator system up to 9 MeV with 125 mA CW deuteron beam will be installed and tested in Rokkasho. Major components of the accelerator are under development or fabrication in EU. The first component of the accelerator, an injector with an ECR ion source, will be delivered to Rokkasho in March 2013.

Numerical techniques for parallel dynamics in electromagnetic gyrokinetic simulations are introduced to regulate unphysical grid-size oscillations in the field-aligned coordinate. It is found that a fixed boundary condition and the nonlinear mode coupling in the field-aligned coordinate, as well as numerical errors of non-dissipative finite difference methods, produce fluctuations with high parallel wave numbers. The theoretical and numerical analyses demonstrate that an outflow boundary condition and a low-pass filter efficiently remove the numerical oscillations, providing small but acceptable errors of the entropy variables. The new method is advantageous for quantitative evaluation of the entropy balance that is required for obtaining a steady state in gyrokinetic turbulence.

In the Broader Approach framework, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, the International Fusion Energy Research Center (IFERC) project, and the Satellite Tokamak project are implemented. In the IFMIF/EVEDA project, engineering design of IFMIF and engineering R&D include the construction and tests of an IFMIF prototype accelerator system up with a 9 MeV and ON deuteron beam, a liquid lithium test loop with free surface flow, and full scale irradiation test module including temperature control instrumentation. The commissioning of the EVEDA lithium test loop was completed in March 2011, and a lithium flow of similar to 5 m/s was obtained. As a part of the IFERC project, R&Ds on reduced activation ferritic/martensitic steels as blanket structural material, SiCf/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out. At the beginning of 2011, the integrated DEMO design team was established among the IFERC project team and EU/JA home teams, where the design criteria, other design basis are discussed as an initial work. A high performance supercomputer with the peak performance of 1.3 Pflops is under installation at the Rokkasho BA site. (C) 2012 Elsevier B.V. All rights reserved.

A hybrid method of semi-Lagrangian and additive semi-implicit Runge–Kutta schemes is developed for gyrokinetic Vlasov simulations in a flux tube geometry. The time-integration scheme is free from the Courant–Friedrichs–Lewy condition for the linear advection terms in the gyrokinetic equation. The new method is applied to simulations of the ion-temperature-gradient instability in fusion plasmas confined by helical magnetic fields, where the parallel advection term severely restricts the time step size for explicit Eulerian schemes. Linear and nonlinear results show good agreements with those obtained by using the explicit Runge–Kutta–Gill scheme, while the new method substantially reduces the computational cost.

A semi-Lagrangian scheme is applied for the first time to computations of charged particle motions along magnetic field lines, to numerically solve the δf gyrokinetic equations in a flux tube geometry. This new solver adopted in the gyrokinetic Vlasov simulations has an advantage over the conventional Eulerian codes in calculating the parallel dynamics, because semi-Lagrangian schemes are free of the Courant-Friedrichs-Lewy (CFL) condition that restricts the time step size. A study of the accuracy of the parallel motion simulations reveals that numerical errors mainly stem from spatial (not temporal) discretization for realistic values of the grid spacing and time step, and it demonstrates the advantage of the semi-Lagrangian scheme. This novel numerical method is successfully applied to linear gyrokinetic simulations of the ion temperature gradient instability, where time steps larger than those restricted by the CFL condition can be employed.

Plan of ITER remote experimentation center (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is described. Objectives of REC activity are (1) to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, (2) to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, and (3) to test the functions of REC and demonstrate the total system by using JT-60SA and existing other facilities in EU. Preliminary identified items to be developed are (1) Functions of the remote experiment system, such as setting of experiment parameters, shot scheduling, real time data streaming, communication by video-conference between the remote-site and on-site, (2) Effective data transfer system that is capable of fast transfer of the huge amount of data between on-site and off-site and the network connecting the REC system, (3) Storage system that can store/access the huge amount of data, including database management, (4) Data analysis software for the data viewing of the diagnostic data on the storage system, (5) Numerical simulation for preparation and estimation of the shot performance and the analysis of the plasma shot. Detailed specifications of the above items will be discussed and the system will be made in these four years in collaboration with tokamak facilities of JT-60SA and EU tokamak, experts of informatics, activities of plasma simulation and ITER. Finally, the function of REC will be tested and the total system will be demonstrated by the middle of 2017.

The confinement of energetic ions generated by neutral beam injection (NBI) and ion cyclotron resonance frequency heating is studied using GNET simulation code, in which the drift kinetic equation ...

Theoretical MHD analyses of the Large Helical Device (LHD) plasmas have progressed extensively. Recent results with respect to three-dimensional equilibrium, linear stability, and nonlinear dynamic...

Heliotron reactors have several features suitable for a fusion power plant, such as no need for current drive, no plasma current disruptions, suitability for steady-state operation, and a wide spac...

Energetic ion-driven magnetohydrodynamic instabilities such as Alfven eigenmodes (AEs), energetic particle modes (EPMs), and their impacts on energetic ion confinement are being studied on the Larg...

In the vacuum of the Large Helical Device (LHD), we can change the plasma volume, the aspect ratio, the ellipticity, the rotational transform, and the height of the magnetic hill through th...

This paper reports the Monte-Carlo calculation of the neoclassical diffusion and viscosity coefficients with the incompressible E × B drift. A recently developed Monte-Carlo code, based on the Einstein-Helfand relation, is extended for this purpose. Present status of a Monte-Carlo-based moment transport code to evaluate the neoclassical flows for multiple species plasmas is also discussed (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The numerical method for analysis of the “MHD stability beta limit” based on a hierarchy integrated simulation code TASK3D has been developed. The numerical model for the effect of the MHD instabilities is introduced such that the pressure profile is flattened around the rational surface due to the MHD instabilities. The width of the flattening of the pressure gradient is determined from the width of the eigenmode structure of the MHD instabilities (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

A numerical modelling of the dynamics of an edge-localized mode (ELM) crash in the spherical tokamak is proposed by means of a three-dimensional nonlinear magnetohydrodynamic simulation. The simulation result shows a crash of the edge pressure profile due to the spontaneous growth of the ballooning mode instability. The simulation result shows good agreement in several characteristic features of the experimental observation of large scale ELMs in an appropriate time scale: (1) relation to the ballooning instability, (2) intermediate-n precursors, (3) low-n structure on the crash, (4) formation and separation of the filament and (5) considerable amount of convective loss of plasma, where n is the toroidal mode number. Furthermore, the model is verified by examining the effect of diamagnetic stabilization and by comparing the nonlinear behaviour with that of the peeling modes. The ion diamagnetic drift terms are found to stabilize some specific components linearly; nevertheless they are not so effective in the nonlinear dynamics such as the filament formation and the amount of loss. For the peeling mode case, no prominent filament structure is found to appear in contrast to the ballooning mode case.

Direct numerical simulations in fully three-dimensional geometry have been performed in order to investigate effects of both the parallel flow to the magnetic field lines discarded in the reduced equations and the parallel thermal conductivity on the nonlinear evolution of interchange instability in a low-β magnetohydrodynamic equilibrium in the inward-shifted vacuum configuration of the Large Helical Device. As the parallel thermal conductivity becomes larger, the perpendicular velocity of the linear eigenfunction becomes relatively smaller than the parallel velocity, and in the nonlinear phase, the energy of the perpendicular velocity is selectively damped by the viscosity. Consequently, the plasma is finally saturated with flattened pressure profile, finite parallel velocity, and fairly good magnetic flux surfaces.

In the three-dimensional magnetic confinement configurations, the results of local mode analyses of the ballooning modes in the covering space (quasi modes) cannot be directly connected by superposition to the global mode analyses of the ballooning modes in the configuration space (physical modes) because of the lack of symmetry. However, a qualitative relation has been established to connect the quasi modes to physical modes in planar axis heliotron configurations with a large Shafranov shift. This relation is based on the topological structure of the level surfaces of the eigenvalues of the quasi modes. High-beta magnetohydrodynamic equilibria in the inward-shifted Large Helical Device configuration are examined. It is shown that the core plasma stays in the second stability, and the peripheral plasma stays near the marginally stable state against ballooning modes.

Local transport properties of high-beta plasmas in the Large Helical Device are studied by comparing the beta dependence of the experimental results with that of the gyro-reduced Bohm-type transpor...

Nonlinear behavior of the resistive wall modes (RWMs) has been studied using nonlinear three-dimensional simulation code based on the reduced magnetohydrodynamic equations in a cylindrical tokamak. The vacuum is replaced by a highly resistive plasma using the pseudo-vacuum model. For the case that the KWM with (m, n) = (3, 1) mode and the internal (m, n)=(5, 2) mode are unstable, the RWM with (m, n) = (3, 1) mode enhances the growth rate of the internal (m, n) = (5, 2) mode in the nonlinear phase. The magnetic field line stochastization is obtained around the plasma cdge due to the overlapping of the magnetic islands of the (m, n) = (3, 1) mode and the (m, n) = (5, 2) mode.

Direct numerical simulation of fully three-dimensional, compressible and nonlinear magnetohydrodynamic equations in the Large Helical Device is carried out in combination with the passive particle simulation. In the simulation. strong vortical motions are excited by the pressure-driven instability and form the mushroom-like structures of pressure. It is shown by the passive particles analysis that the fluid volumes around the resonant magnetic surfaces experience finite ronipressibility and toroidal deformation, which are both excited by the strong vortical motions. The passive particles simulation helps us to investigate local structures even for low Fourier wavenumber modes.

In local island (m/n = 1/1) divertor discharges in the large helical device a stable super dense core plasma develops when a series of pellets are injected. A core region with a density as high as 4.6 × 1020 m−3 and a temperature of 0.85 keV is maintained by an internal diffusion barrier with a very high density gradient. In a study of island dynamics, we find that an externally imposed large island (m/n = 1/1) as large as 15% of the minor radius is healed when beta at the island exceeds a critical value.

Development of an integrated simulation system for helical plasma is described that draws new experimental plans, including those in new devices, and does experimental data analysis from the viewpoint of integrated physics. The integrated simulation system to be developed has a modular structure that consists ofmodules for calculating magnetohydrodynamic (MHD) equilibrium/ stability, transport, and heating. Each module can be selected in accordance with a user's request and can be combined with other modules. When we want to perform the integrated simulation during the entire plasma duration, a transport module is to be a core module. An integrated tokamak transport code will be extended for the helical configuration and used as a transport module. As the first step of the extension, time evolution of the plasma net current, which is consistent with the three-dimensional MHD equilibrium, is planned to be solved for Large Helical Device plasmas by taking into account the bootstrap current and the beam-driven current.

The three-dimensional (3D) magnetohydrodynamic (MHD) equilibrium solver HINT is improved as a new HINT code, 'HINT2' having various useful features by introducing modern computing techniques. By using the algorithm inherent to HINT2, the treatment of the peripheral plasma corresponding to the experimental results is addressed in the Large Helical Device (LHD).

The motion of an ablation cloud is investigated in a tokamak by using ideal magnetohydrodynamic equations with ablation processes. The cloud quickly expands along the magnetic field and simultaneously drifts to the lower field side due to a tire tube force and 1/R force. It is found that the acceleration of the cloud increases and subsequently decreases due to the energy deposit from the bulk plasma and, finally, the cloud is stopped due to the poloidal field.

In order to investigate the separatrix structure inside the plasma, the magnetohydrodynamic (MHD) equilibrium and the high-energy particle orbit are studied for the Large Helical Device plasmas with two magnetic axes located within the two lobes of a figure-eight magnetic separatrix. Magnetic axes are split for strongly elongated cross section by the external quadrupole field. MHD equilibria without assuming existence of nested flux surfaces are obtained from the HINT code. For the vertical elongation, the figure-eight structure evolves due to the finite beta effect and the structure of magnetic field lines outside the separatrix is ergodized. For the horizontal elongation, the Shafranov shift is very large Δ/a > 0.5. To see MHD equilibrium beta limit, high beta equilibrium (~5%) is studied. The magnetic axis shifts extremely to the outside of the torus and a new figure-eight separatrix appears. The particle orbit on the separatrix is studied by solving the guiding-centre drift equation on the rectangular grid. If the reflecting point of the trapped particle is close to the separatrix, the banana width is very large implying large, nonlocal neoclassical transport.

Neoclassical transport simulation code (FORTEC-3D) applicable to both axisymmetric and non-axisymmetric configurations is developed to investigate non-local effects on neoclassical transport phenomena. The time evolution of the radial electric field is simulated in the full volume of the confinement region of tokamak and helical model plasmas. It is found that the damping rate of the geodesic-acoustic-mode (GAM) oscillation becomes faster than that predicted from a single-surface transport analysis. The time evolution of the radial electric field towards the ambipolar state shows a non-local behaviour, which indicates a coupling of GAM oscillation between the neighbouring two flux surfaces because of the finite-orbit-width effect.

Pressure driven, ideal ballooning stability calculations are often used to predict the achievable plasma β in stellarator configurations. In this paper, the sensitivity of ballooning stability to plasmas profile variations is addressed. A simple, semi-analytic method for expressing the ballooning growth rate, for each field line, as a polynomial function of the variations in the pressure-gradient and the average magnetic shear from an original equilibrium has recently been introduced (Hudson and Hegna 2004 Phys. Plasmas 11 L53). This paper will apply the expression to various stellarator configurations and comment on the validity of various truncated forms of the polynomial expression. In particular, it is shown that it is generally insufficient to consider only the second order terms as previously assumed, and that higher order terms must be included to obtain accurate predictions of stability.

The orbit topology and confinement of neutral beam-injected energetic ions are investigated for the current target configuration of the CHS-qa quasi-axisymmetric stellarator. It was shown that tangentially co-injected neutral beam (NB) heating is efficient even at a low magnetic field strength Bt of 0.5 T, whereas the heating efficiency of the counter-injected NB becomes significantly lower as Bt decreases because of the increase of first orbit loss. The energy loss rate increases as the beam injection angle becomes perpendicular, suggesting that the residual non-axisymmetric ripple in the peripheral domain plays a role in enhancing the transport of trapped ions. An interesting observation involves the appearance of the island structure in both the gyro motion following orbit and the guiding center collisionless orbit of counter-moving transit beam ions. It appears under a particular, narrow range of parameters, i.e., energy, pitch angle v///v, normalized minor radius r/a at the launching point and Bt.

The global ideal magnetohydrodynamic (MHD) stability for a proposed compact quasi-axisymmetric stellarator CHS-qa has been investigated taking the effect of bootstrap current into account. Assuming experimentally achievable density and temperature profiles, the stability properties of global low-n modes have been studied by using threedimensionalnumerical codes based on fixed boundary MHD equilibria including self-consistent bootstrap current for the CHS-qa reference configuration. Consequently it has been shown that values of edge rotational transform play a crucial role in triggering external kink instability. Concerning a lot of other possibilities in experimental practice to change the total parallel current, we have also studied equilibria with increased or decreased parallel current, but fixed profile. The onset of external kink modes depends on rotational transform or current profile, and we found a stable equilibrium in spite of the edge rotational transform above 0.5. The results imply the possibility of stabilizing external kink modes through current and/or pressure profile control in high beta equilibria.

Magnetohydrodynamic (MHD) equilibria of Heliotron J plasma are investigated by using the HINT code. HINT calculates an MHD equilibrium by using the relaxation method, which calculates the time evolution of dissipative MHD equations. Since HINT uses Eulerian rectangular grids, it does not assume nested flux surfaces. In the standard configuration, magnetic islands appear inside the plasma in spite of low {beta} ( {approx} 0.4%). The width of the islands depends on the pressure distribution. To reduce island width, feedback control of the external vertical field is introduced. Because of the effect of the external vertical field, magnetic islands are suppressed. The effects of the net toroidal currents on MHD equilibrium are also investigated. The rotational transform is changed by the currents, but the Shafranov shift is almost unaffected. The width of the islands changes because of the change in the rotational transform.

A two-dimensional hydrodynamic simulation code CAP has been developed in order to investigate the dynamics of hydrogenic pellet ablation in magnetized plasmas throughout their temporal evolution. One of the properties of the code is that it treats the solid-to-gas phase change at the pellet surface without imposing artificial boundary conditions there, as done in previous ablation models. The simulation includes multispecies atomic processes, mainly molecular dissociation and thermal ionization in the ablation flow beyond the pellet, with a kinetic heat flux model. It was found that ionization causes the formation of a quasistationary shock front in the supersonic region of the ablation flow, followed by a “second” sonic surface farther out. Anisotropic heating, due to the directionality of the magnetic field, contributes to a nonuniform ablation (recoil) pressure distribution over the pellet surface. Since the shear stress can exceed the yield strength of the solid for a sufficiently high heat flux, the solid pellet can be fluidized and flattened into a “pancake” shape while the pellet is ablating and losing mass. The effect of pellet deformation can shorten the pellet lifetime almost 3× from that assuming the pellet remains rigid and stationary during ablation.

The nonlinear behavior of the interchange mode in Large Helical Device (LHD) plasma depends on the overlap of the vortices with different helicity. If the vortices are separated in the radial direction, each mode saturates mildly with generating the local flat regions in the pressure profile. In the case of significant overlap of the vortices, the appearance of large-scale convection results in sudden global reduction of the pressure. The sequence of the saturated pressure profile in the increase of beta can suppress the overlap. Self-organization of the pressure profile to suppress the overlap of the vortices can be the stabilizing mechanism in the LHD plasma.

A low aspect ratio stellarator CHS-qa with a quasi-axisymmetric configuration has been designed at NIFS, whose main objective is the experimental study of confinement improvement. The high beta equilibrium was obtained with the HINT code which does not require the assumption of nested magnetic surfaces. The equilibria with bootstrap current were calculated using the SPBSC code and their global MHD stability was evaluated with the TERPSICHORE code. The boundary structure of the magnetic field was examined to scope divertor options. A new method of calculating the neoclassical transport properties was developed for toroidal configurations with very small non-axisymmetry. The control knobs of the configuration properties are proposed for the study of improved confinement.

Free boundary MHD equilibrium properties of Heliotron J are investigated by VMEC, HINT and PIES codes, and ideal MHD stability properties are studied by the Mercier criterion, the ballooning mode equation and the CAS3D global stability code. It is shown by the equilibrium calculations that the change of the plasma boundary shape is substantial in a low shear helical system even if β is relatively low. Preliminary comparison between PIES results and HINT results shows that the β value at which the magnetic island begins to be perceptible is almost the same in both codes, but the island width seems to be different. It is considered that the difference comes from a difference of pressure distribution in real space which is determined through each numerical algorithm. In the stability analysis, the effect of the global magnetic shear on the three-dimensional or helical ballooning mode whose mode structure shows strong poloidal and toroidal mode (helical mode) coupling is investigated. It is found that the positive shear of the rotational transform is favourable for the three-dimensional ballooning mode stability in a low shear helical system.

We analyze the relationship between the experimentally observed pressure gradients at resonant rational surfaces and the theoretically predicted ideal magnetohydrodynamics (MHD) unstable region of global modes in the large helical device (LHD). According to the stability analysis of the ideal MHD modes with a low toroidal mode number, we find that the ideal MHD mode gives a constraint on the operational regime of the pressure gradients in the core. In the edge, a clear saturation of the pressure gradients due to the ideal MHD instability has not been observed up to the high beta regime around 3% as the volume-averaged toridal beta value, where global ideal MHD modes are predictedto be unstable.

For the large helical device (LHD), the nonlinear evolution of equilibria that are linearly unstable to ideal interchange modes is studied using the reduced MHD equations. At sufficiently low beta, each individual mode saturates without affecting directly the evolution of the other modes. They only couple through the modification of the averaged pressure profile. The change of the averaged pressure profile is limited to the local flattening near the resonant surfaces. At higher beta values and for the same initial pressure profile, a bursting phenomenon in the kinetic energy is observed. This bursting activity is caused by the overlap of multiple modes, which results in a global reduction of the pressure. However, increasing beta and using a pressure profile obtained from the nonlinear evolution at the lower beta suppress this bursting behaviour. This result indicates that the pressure profile can be self-organized so that the LHD plasma could reach a high beta regime through a stable path.