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Magnetic interaction between a tokamak reactor and its reinforced-concrete building has been analyzed using the analytical method and ANSYS electromagnetic code. Although it is better to use the non-magnetized material for the foundation of the air core tokamak building, it is interesting to study whether the magnetic material can be used in order to save the total cost of a fusion reactor. This study is motivated from the present ITER under construction in France. Assuming the iron plate approximation instead of discrete rebars, we analyzed the effect of the magnetized material on the magnetic field null regime in the plasma break down phase, x point location in the divertor operation, the magnetic force applied to the floor, and the induction effect on the iron floor. Although the effect of the magnetic material on the plasma performance is found to be not so crucial, the mechanical soundness of the floor due to the magnetic attractive force is most concern for machine safety.

With respect to a particle fueling method in a spherical tokamak reactor with an advanced fuel, we investigate the availability of a merging process following the translation of a small torus plasma through which a plasma current flows in the same direction as the main plasma. A three-dimensional resistive MHD model is used to examine the particle fueling effect by reproducing the translation process and simultaneously performing an ion trajectory calculation. The simulation shows the merging process between the translated plasma and the main plasma and exhibits potential as a particle fueling method. The results of the study indicate that up to 30% of the ions in the secondary plasma are supplied to the main plasma.

The coronavirus disease 2019 (COVID-19) has been damaging our daily life after declaration of pandemic. Therefore, we have started studying on the characteristics of Susceptible-Infectious-Recovered (SIR) model to know about the truth of infectious disease and our future.After detailed studies on the characteristics of the SIR model for the various parameter dependencies with respect to such as the outing restriction (lockdown) ratio and vaccination rate, we have finally noticed that the second term (isolation term) in the differential equation of the number of the infected is quite similar to the “helium ash particle loss term” in deuterium-tritium (D-T) nuclear fusion. Based on this analogy, we have found that isolation of the infected is not actively controlled in the SIR model. Then we introduce the isolation control time parameter q and have studied its effect on this pandemic. Required isolation time to terminate the COVID-19 can be estimated by this proposed method.To show this isolation control effect, we choose Tokyo for the model calculation because of high population density. We determine the reproduction number and the isolation ratio in the initial uncontrolled phase, and then the future number of the infected is estimated under various conditions. If the confirmed case can be isolated in 3∼8 days by widely performed testing, this pandemic could be suppressed without awaiting vaccination. If the mild outing restriction and vaccination are taken together, the isolation control time can be longer. We consider this isolation time control might be the only solution to overcome the pandemic when vaccine is not available.

In the present RF-driven (ECCD) steady-state plasma on QUEST (Bt = 0.25 T, R = 0.68 m, a = 0.40 m), plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to plasma current fitting (PCF) method. We consider that the current in the open magnetic surface is due to orbit-driven current by high-energy particles in RF-driven plasma. So based on the analysis of current density profile based on the orbit-driven current, plasma equilibrium is to be calculated. We calculated high energy particles guiding center orbits as a contour plot of conserved variable in Hamiltonian formulation and considered particles initial position with different levels of energy and pitch angles that satisfy resonance condition. Then the profile of orbit-driven current is estimated by multiplying the particle density on the resonance surface and the velocity on the orbits. This analysis shows negative current near the magnetic axis and hollow current profile is expected even if pressure driven current is considered. Considering the hollow current profile shifted toward the low-field region, the equilibrium is fitted by J-EFIT coded by MATLAB.

As QUEST has a small central solenoid (CS), a larger Ohmic discharge current has been obtained when the plasma shifts to the inboard side. This tendency restricts a divertor operation to the smaller plasma current regime. As the inner divertor coil has a smaller mutual inductance, it would be expected that its utilization seems to be better for easier plasma current ramp-up for a divertor operation. In this work, we made comparative studies on the plasma current ramp-up for two divertor coils. It is found that while the inner divertor coil with smaller mutual inductance needs a larger coil current, the outer divertor coil with larger mutual inductance needs a smaller coil current for divertor operation. Thus we have found that the plasma current ramp-up characteristics are almost similar for both configurations. We also propose a new fueling method for spherical tokamak (ST) using the coaxial helicity injection (CHI). The main plasma current would be generated at first, and then the CHI plasma current is created between bottom two electrode plates and merged into the main plasma current for fueling.

In the present RF-driven divertor plasma of QUEST, it has been observed that orbit-driven current flows in the open magnetic surfaces outside of the closed magnetic surfaces. To observe this phenomenon and the characteristics of the orbit-driven current, current density profiles have been calculated on two different equilibrium conditions. We calculated current density profiles from particle guiding center orbits both for the fundamental and the second harmonic resonances for the 8.2 GHz electron cyclotron current drive. From this calculation, hollow current density profiles have been obtained with significant characteristics on both conditions. Only positive current distribution has been observed in the open magnetic surfaces outside of the closed magnetic surfaces.

In the present RF-driven (ECCD) steady-state plasma on QUEST, the plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to the plasma current fitting (PCF) method. The current in the open magnetic surface is due to the orbit-driven current by high-energy particles in the RF-driven plasma. High-energy particles guiding center orbits are calculated as contour plots of conserved variables in the Hamiltonian formulation considering particles in the initial position with different energies and pitch angles satisfying the resonance condition. A negative current appears near the magnetic axis, and a hollow current profile is expected even if the pressure driven current is considered. The equilibrium is fitted within nested magnetic surfaces by J-EFIT coded by MATLAB using the hollow current profile shift toward the low-field region. Although the plasma boundary shape reflects the plasma current density profile, the equilibrium shape fitted by J-EFIT does not coincide with the orbit-driven current profile. However, introducing an extension of the current profile without assuming nested contours into the J-EFIT code appropriately fits the plasma shape with the hollow current profile to the measured magnetic data.

Fully non-inductive plasma start-up was successfully achieved by using a well-controlled microwave source on the spherical tokamak, QUEST. Non-inductive plasmas were maintained for approximately 3–5 min, during which time power balance estimates could be achieved by monitoring wall and cooling-water temperatures. Approximately 70%–90% of the injected power could be accounted for by calorimetric measurements and approximately half of the injected power was found to be deposited on the vessel wall, which is slightly dependent on the magnetic configuration. The power distribution to water-cooled limiters, which are expected to be exposed to local heat loads, depends significantly on the magnetic configuration, however some of the deposited power is due to energetic electrons, which have large poloidal orbits and are likely to be deposited on the plasma facing components.

In order to evaluate the magnetic surface current density distributions, the electron guiding center orbits were calculated with the equilibrium fitting plasma equilibrium analysis. The current density distributions in electron cyclotron-heated plasma were estimated from the orbit analyses in the equilibrium plasma shaping for the fundamental and the second-harmonic resonant electrons separately. The current density distribution profiles on the equatorial plane were obtained for the electrons with initially positive and negative velocities parallel to the magnetic field direction as well as the trapped electrons. The surface averaged current density profiles of the closed flux surfaces were also evaluated. A significant amount of the positive current density distributions appeared outside the last closed flux surface (LCFS), while all the negative current density distributions were evaluated inside the LCFS. The trapped electrons being the second-harmonic electron cyclotron resonant contributed the negative current inside the LCFS. The positive current density distributions outside the LCFS were significant for the driven current density for both the fundamental and second-harmonic resonances.

In the Force Free Helical Reactor FFHR-d1 (R∼15.7 m, a∼2.5 m, Bo ∼4.5 T, 〈β〉∼5 % and the fusion power of 3 GW) [1, 2], it is demonstrated that the thermally unstable operation is better to achieve the higher density and lower temperature plasma (n(0)∼ 9 × 1020 m−3 and T(0)∼7 keV) than the impurity injection method. One drawback of the thermally unstable operation is a possibility of the thermal runaway when fueling systems have failures. However, we have found that the inherently safe function exists owing to the plasma outward shift during the thermal runaway. Thus we continue to pursue this high-density operation scenario for FFHR-d1. Preliminary estimation of the alpha energy loss fraction of ∼3 % is obtained for a broad density profile in the high-density ignition regime.

After several experimental campaigns in the Kyushu University Experiment with Steady-state Spherical Tokamak (QUEST), the originally stainless steel plasma-facing wall (PFW) becomes completely covered with a deposited film composed of mixture materials, such as iron, chromium, carbon, and tungsten. In this work, an innovative colorimetry-based method was developed to measure the thickness of the deposited film on the actual QUEST wall. Because the optical constants of the deposited film on the PFW were position-dependent and the extinction coefficient k1 was about 1.0–2.0, which made the probing light not penetrate through some thick deposited films, the colorimetry method developed can only provide a rough value range of thickness of the metal-containing film deposited on the actual PFW in QUEST. However, the use of colorimetry is of great benefit to large-area inspections and to radioactive materials in future fusion devices that will be strictly prohibited from being taken out of the limited area.

Plasma operation control of the Large Helical Device (LHD)-type helical fusion reactor, FFHR-d1, was examined by using a quasi-one-dimensional particle balance calculation code. It was found that feedback control of the fusion power, which is proposed in the past study with a zero-dimensional model, was difficult because of the time delay between the pellet injection and response of the fusion power due to shallow pellet deposition. While the achievable parameter range of the fusion power at a steady state is limited by the pellet injection condition and magnetic field strength, smooth ignition access and steady-state sustainment can in principle be achieved via feedback control of the line-averaged electron density by manipulating the pellet fuelling rate and maintaining adequate control of the external heating power.

We propose to use of a small iron core transformer to start up the plasma current in a spherical tokamak (ST) reactor without central solenoid (CS). Taking advantage of the high aspect ratio of the STOR-M iron core tokamak, we have demonstrated that the plasma current up to 10–15 kA can be started up using the outer Ohmic heating (OH) coils without CS, and that the plasma current can be maintained further by increasing the outer OH coil current during iron core saturation phase. When the magnetizing current reaches 1.2 kA and the iron core becomes saturated, the third capacitor bank connected to the outer OH coils is discharged to maintain the plasma current. The plasma current is slightly increased and maintained for additional 5 ms as expected from numerical calculations. Core saturation has been clearly observed on the hysteresis curve. This is the first experimental demonstration of the feasibility of slow transition from the iron core to air core transformer phase without CS. The results implies that a plasma current can be initiated by a small iron core and could be ramped up by additional heating and vertical field after iron core saturation in future STs without CS.

In the present RF-driven plasma with a lot of high-energy electrons, there may be anisotropic plasma pressure, which makes the usual equilibrium analysis difficult, but the Cauchy condition surface method can reconstruct the plasma shape precisely regardless of the anisotropy. In addition, the plasma current effect in the open magnetic surfaces outside of the closed magnetic surfaces is considered in the RF-driven divertor plasma. In the reconstruction process, singular value (SV) decomposition is used and optimal criterion function for generalized cross validation is estimated concerning truncation or reduction of the small-SV components.

Global particle balance in non-inductive long-duration plasma on QUEST has been investigated. Approximately 70% of the fuel hydrogen (H) was retained in the wall and then was almost exhausted just after the discharge. The global recycling ratio (Rg), defined as the ratio of the evacuated H2 flux to that injected, was found to gradually increase during discharges and subsequently rose rapidly. To study the growth of Rg, the thermal desorption spectra after deuterium implantation in a specimen exposed to QUEST plasma was analyzed with a model which includes reflection, diffusion, solution, recombination, trapping, and plasma-induced desorption in the re-deposition layer. The model reconstructs the growth of Rg during a long-duration plasma and indicates solution plays a dominant role in the growth.

We present the newly discovered control algorithm for the heating power during ignition access to the thermally unstable ignition regime in the FFHR helical reactor. We used the fact that the thermally unstable and stable ignition regimes on Plasma Operating Contour map (POPCON) have an inverse relationship with respect to the density and temperature. As the density limit restricts the operating point to the thermally stable regime, the postulated temperature limit restricts the operating point to the thermally unstable regime. As a result, the heating power for feedback control is expressed by the measured temperature and power balance terms such as alpha heating and bremsstrahlung loss powers. The external heating power thus obtained is applied during the low-density NBI penetration phase, and then switched off automatically after over-passing the saddle point by alpha-heating power, reaching high-density ignition regime. This control algorithm is robust to various perturbations of the confinement factor, impurity content and so on. Feedback control algorithms thus established for both fueling and heating makes the thermally unstable operation more realistic in FFHR.

A plasma control system to sustain divertor configurations is developed on QUEST (Q-shu university experiment with steady-state spherical tokamak). Magnetic fluxes are numerically integrated at 100 kHz using FPGA (Field-Programmable Gate Array) modules and transferred to a main calculation loop at 4 kHz. With these signals, plasma shapes are identified in real time at 2 kHz under the assumption that the plasma current can be represented as one filament current. This calculation is done in another calculation loop in parallel by taking advantage of a multi-core processor of the plasma control system. The inside and outside plasma edge positions are controlled to their target positions using PID (proportional-integral-derivative) control loops. Whereas the outside edge position can not be controlled by the outer PF coil current, the inside edge position can be controlled by the inner PF coil current.

QUEST has a divertor configuration with a high and a negative n-index, and the problem of plasma vertical position instability control in QUEST is still under extensive study for achieving high efficiency plasma. The instability we considered is that the toroidal plasma moves either up or down in the vacuum chamber until it meets the vessel wall and is extinguished. The actively controlled coils (HCU and HCL) outside the vacuum vessel are serially connected in feedback with a measurement of the plasma vertical position to provide stabilizing control. In this work, a robust controller is employed by using the loop synthesis method, and provides robust stability over a wide range of n-index. Moreover, the gain of the robust controller is lower than that of a typical proportional derivative (PD) controller in the operational frequency range; it indicates that the robust controller needs less power consumption than the PD controller does.

A CW phased-array antenna system for electron cyclotron/Bernstein wave heating and current drive (ECWH/CD, EBWH/CD) experiments was developed in the QUEST. The antenna was designed to excite an elliptically polarized pure O-mode wave in oblique injection for the O-X-B mode conversion scenario, and its good performance was confirmed at a high power level. Long pulse discharges with a plasma current of 10 kA and 15 kA were non-inductively attained for 37 s and 20 s, respectively, with only radio frequency (RF) power. Divertor configurationswere also obtained in the RF-sustained plasmas. A new operational window for sustained plasma current was observed in the high-density plasma with a higher RF incident power. Two new heating and current drive systems with an 8.56GHz klystron and a 28GHz gyrotron are being prepared to conduct CW EBWH/CD experiments in the high-density plasma. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

Based on high-density and high-temperature plasma experiments in the large helical device (LHD), conceptual design studies of the LHD-type helical DEMO reactor FFHR-d1 have been conducted by integrating wide-ranged R&D activities on core plasmas and reactor technologies through cooperative researches under the fusion engineering research project, which has been launched newly in NIFS. Current activities for the FFHR-d1 in this project are presented on design window analyses with designs on core plasma, neutronics for liquid blankets, continuous helical magnets, pellet fueling, tritium systems and plasma heating devices.

CCS (Cauchy Condition Surface) method is a numerical approach to reproduce plasma shape, which has good precision in conventional tokamaks. In order to apply it in plasma shape reproduction of ST (Spherical Tokamak), the calculation precision of the CCS method in a spherical tokamak CPD (Compact PWI experimental Device) (Bt = 0.25 T, R = 0.3 m, a = 0.2 m) has been analyzed. The precision was confirmed also in ST and decided to be applied to a spherical tokamak QUEST (Bt = 0.25 T, R = 0.68 m, a = 0.40 m).. In present stage from the magnetic measurement, it is known that the eddy current effect is large in QUEST experiment, and there are no special magnetic measurements for eddy current now, so some proper model should be selected to evaluate the eddy current effect. The eddy current density by not only CS (Center Solenoid) coil but also plasma current is calculated using EDDYCAL (JAEA). The eddy current magnitudes are taken as unknown variables and solved together with plasma shape reconstruction in ohmic discharge and ECCD (Electron Cyclotron Current Drive) discharge.

A scenario to verify the knock-on tail formation by energetic protons and the resulting enhancement of the fractional ion heating by using 3He-containing deuterium plasma is presented. On the basis of the Boltzmann-Fokker-Planck (BFP) model, the knock-on tail formation in deuteron distribution function due to nuclear plus interference (NI) scattering of fusion-produced energetic protons and the resulting modification of the neutron emission spectrum are examined. A recognizable change in the energetic-neutron emission rate due to the knock-on tail formation depending on the external 3He inclusion is shown. Another possible scenario to ascertain the enhancement of ion-heating by using γ-ray-generating 6Li+D reaction is also presented and discussed.

Both start-up and sustainment of plasma were successfully achieved by fully non-inductive current drive using microwave with a frequency of 8.2 GHz. Plasmas current of 15 kA was implemented for 1 s. Magnetic surface reconstruction exhibited a plasma shape with an aspect ratio of below 1.5. The plasma current was dependent significantly on the launched microwave power and vertical magnetic field, while not affected by the mode of launched wave and the toroidal refractive index. Hard X-ray (HXR) emitted from energetic electrons accelerated by the microwave was observed, and the discharge with a plasma current over 4 kA followed the same trend as the number of photons of 10 keV to 12 keV. This suggests that the plasma current may be driven by energetic electrons. Based on the experimental conditions, alternative explanations of how the plasma current could be driven are discussed.

Particle retention and recycling in plasma fusion devices are generally associated with the diffusion of atomic hydrogen into the materials. The resulted permeation of atomic hydrogen is known as plasma driven permeation (PDP). This permeation may also be significant, even in the walls, which are not directly exposed to the plasma. Under similar conditions, the permeation flux (Γperm) of hydrogen through a 30 μm thick Ni membrane heated at 412–575 K has been measured in the spherical tokamak QUEST. Γperm is being measured during the scans of different operating parameters like RF power (PRF), chamber pressure (Pchamber), discharge widths (τdis) and vertical magnetic field (BZ). Simultaneously edge plasma density and spectral intensities of atomic (Balmer) lines and molecular (Fulcher) bands have been compared with the permeation measurements. A linear relationship has been established between the time integrated Γperm i.e. permeation fluence (Qperm) and the time integrated Hα intensity i.e. Hα fluence (Qα). Qperm also shows a strong relationship with the edge plasma density and various spectral fluences. The obtained results are discussed for exploring the applicability of the permeation probes in measuring the atomic flux near the first walls.

Initial results from the recently implemented transient coaxial helicity injection (CHI) system on QUEST are reported. QUEST uses a new CHI electrode configuration in which the CHI insulator is not part of the vacuum boundary, making this configuration easier to implement in fusion reactors. Experimental results show that transient CHI startup in this alternate electrode configuration is indeed possible. Reliable gas breakdown was achieved, and toroidal currents up to 45 kA were generated.

Three-dimensional resistive MHD simulation was carried out in order to develop a method to fuel the plasma core by merging two Spherical Torus (ST) plasmas. A relatively small ST plasma was translated into the axial direction by external magnetic field control and made to collide with the larger ST. In this calculation, ballooning instability occurred in both plasmas during translation, while the magnetic surface was observed to collapse from the Poincare plot of the magnetic lines. It was observed that the magnetic lines in the core regions of the two STs were connected to each other and that the merging was completed.

AC Ohmic coil operation experiments with frequencies up to 10 kHz were performed on the TST-2 spherical tokamak device, and the pre-ionization process was studied in detail. The minimum loop voltage for pre-ionization was 0.4 V, which corresponds to 0.5 V ${\rm m}^{-1}$ at the inboard limiter. Dependences of growth rate and saturation level of the process on various parameters were obtained, and they are compared with a time-dependent 0-dimensional model based on Townsend avalanche and loss along field lines. Most of the dependences are reproduced qualitatively by the model, and quantitative differences are within a factor of several. However, the external vertical field dependence of the appearance time, which is defined as the time to observe a plasma, and the isotope effect cannot be reproduced by the model. An ambipolar diffusion state which is predicted theoretically but mitigated experimentally is discussed. It was found that secondary electron emission at the limiter surfaces is a candidate mechanism to mitigate the state.

The various plasma current start-up techniques and related topics in a spherical tokamak (ST) device are described. The Ohmic heating coil current clamp experiments in NSTX are described and discussed, and the plasma current start-up experiments in the STOR-M tokamak with iron core and the outer vertical field coil is presented as one of technique for a plasma current start-up in a ST.

Progress of design studies on a heliotron reactor is reported. Recent experimental achievements in large helical device (LHD) are building up the physical basement. Integrated designs on LHD-type energy reactors have improved the base design of force free helical reactor and revealed economical feasibility. Large helical device is exploring a new horizon of net current free helical plasmas. Inherent characteristics demonstrate reactor relevant performances individually; steady state (1 h), high beta (5%), high density (1 × 10 21 m −3 ), etc. In particular, discovery of internal diffusion barrier provides a novel attractive scenario of super high density operation. It greatly facilitates physical requirements as well as mitigates engineering demands. New methods have been proposed to access and operate ignited plasmas; a long rise-up over 300 s to reduce the heating power to 30 MW and a new proportional-integration-derivative (PID) control of the fueling to handle the thermally unstable plasma at high-density operations.

Minimization of the external heating power to access self-ignition is advantageous to increase the reactor design flexibility and to reduce the capital and operating costs of the plasma heating device in a helical reactor. In this work we have discovered that a larger density limit leads to a smaller value of the required confinement enhancement factor, a lower density limit margin reduces the external heating power and over 300 s of the fusion power rise-up time makes it possible to reach a minimized heating power. While the fusion power rise-up time in a tokamak is limited by the OH transformer flux or the current drive capability, any fusion power rise-up time can be employed in a helical reactor for reducing the thermal stresses of the blanket and shields, because the confinement field is generated by the external helical coils.

A new scenario to control transport of 'cold' ?-particles in a helical fusion device by changing poloidal field coil current during plasma discharge is proposed here. A way to enhance radial transport of ?-particles in the intermediate energy range is considered which relies on specific features of helical magnetic field. To remove helically trapped, cold ?-particles, a ?-induced change in the B/B0 modulation along the particle trajectory is employed. To remove passing particles resonant effects connected with the poloidal field coil currents are used.

We investigate and calculate particle orbits and the effect of particle orbits on plasma current density for nonrelativistic resonance condition in the present RF-driven divertor plasma on QUEST. We surveyed particle orbits for different values of parallel refractive index, particle initial positions and pitch angles on fundamental and second harmonic resonance conditions. We observed that for fundamental harmonic resonance condition when particle orbits are plotted on the poloidal cross-section for positive values of parallel refractive index, these orbits are started from the resonance surface and produced their orbits around the LCFS (Large Closed Flux Surface). These orbits carry positive current. When particle orbits are plotted for negative values of parallel refractive index, these orbits are started from resonance surface, but remained at the inside of the LCFS. These orbits carry negative current that reduced the overall plasma current. For second harmonic resonance condition when particle orbits are plotted on the poloidal cross-section most of the orbits remained in inside the LCFS and carry positive current. When we consider the value of parallel refractive index −0.4 and +0.4 some particle orbits arrived at the limiter and become lost particles. On the other hand, when we consider particle initial positions 0.16 m or more vertically far from the mid plane some banana orbits are produced. These banana orbits make the current density profile maximum at low field side region. From this calculation we got a hollow current density profile with current density peak at the low field side region outside of the LCFS. From this calculation we can infer that parabolic current density profile is possible, if we set the resonance surface outside of the magnetic axis by increasing the toroidal magnetic field coil current and make the plasma position inward by increasing vertical field coil current.

A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. This result suggests a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. The effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments in the STOR-M tokamak.

Ignition capability is studied and compared for two types of D–3He spherical tokamak (ST) and medium aspect ratio tokamak (MT) reactors. The required power fraction to ions for maintaining the hot ion mode, which is indispensable for D–3He fusion, is smaller in a ST reactor than that in a MT reactor due to the lower synchrotron radiation loss. It is found that the nuclear elastic scattering capable of supplying 40% of the fusion energy to ions is enough to maintain the hot ion mode in the D–3He ST reactor with the high wall reflectivity. For achieving the D–3He MT reactor, the power fraction to ions should be additionally increased.

An overview of conceptual design activities on the LHD-type helical reactor FFHR is presented, mainly focusing on optimization studies on the reactor size and the proposal of a long-life blanket. A major radius of around 15 m is the present candidate under the constraints of the energy confinement achieved in LHD, a maximum magnetic field around 13 T with a current density around 30 A/mm2 and a neutron wall loading around 1.5 MW/m2. R&D on super-conducting magnet systems of large scale, high field and high current-density are new challenging targets based on the LHD. The development of new design tools has been started aiming at establishing a virtual power plant (VPP) and a virtual reality system for 3D design assisting. Next design issues are mainly on engineering optimization of the first wall thickness, the detailed 3D blanket system, and unscheduled replacements of breeder blankets.

Several techniques for initiating the plasma current without the use of the central solenoid are being developed in TST-2. While TST-2 was temporarily located at Kyushu University, two types of start-up scenarios were demonstrated. (1) A plasma current of 4 kA was generated and sustained for 0.28 s by either electron cyclotron wave or electron Bernstein wave, without induction. (2) A plasma current of 10 kA was obtained transiently by induction using only outboard poloidal field coils. In the second scenario, it is important to supply sufficient power for ionization (100 kW of EC power was sufficient in this case), since the vertical field during start-up is not adequate to maintain plasma equilibrium. In addition, electron heating experiments using the X–B mode conversion scenario were performed, and a heating efficiency of 60% was observed at a 100 kW RF power level. TST-2 is now located at the Kashiwa Campus of the University of Tokyo. Significant upgrades were made in both magnetic coil power supplies and RF systems, and plasma experiments have restarted. RF power of up to 400 kW is available in the high-harmonic fast wave frequency range around 20 MHz. Four 200 MHz transmitters are now being prepared for plasma current start-up experiments using RF power in the lower-hybrid frequency range. Preparations are in progress for a new plasma merging experiment (UTST) aimed at the formation and sustainment of ultra-high β ST plasmas.

Plasma start-up and sustainment without an inductive field have been studied in the TST-2@K spherical tokamak using high power RF sources (8.2?GHz/up to 170?kW). Steady state discharges with a plasma current of 4?kA were achieved. The line integrated density was about 3 ? 1017?m?2 and the electron temperature was 160?eV. A truncated equilibrium was introduced to reproduce magnetic measurements. It was found that a positive Pfirsch?Schl?ter current in the open field line region at the outboard boundary makes a significant contribution to the current. Insensitivity of the current to variations in the vertical field and RF power variation was also found.

Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Γn ≈ 8.8 × 1013 n s−1 cm−2 ('wall loading' WL ≈ 2 MW m−2), over size-scale >105 cm2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year ('neutron fluence' >0.3 MW yr−1 m−2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~3, Ip ~ 12 MA, BT ~ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.

A broadly based study of the fusion engineering and plasma science conditions of a Component Test Facility (CTF), using the Spherical Torus or Spherical Tokamak (ST) configuration, have been carried out. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes {gamma}{sub n} > 4.4 x 10{sup 13} n/s/cm{sup 2}, over size scales > 10{sup 5} cm{sup 2} and depth scales > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year. The desired chamber conditions can be provided by a CTF with R{sub 0} 1.2 m, A = 1.5, elongation {approx} 3, I{sub p} {approx} 9 MA, B{sub T} {approx} 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. Relatively robust ST plasma conditions are adequate, which have been shown achievable without active feedback manipulation of the MHD modes. The ST CTF will test the single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTFmore » has been prepared and utilized as part of this study. The results show high potential for a family of lower-cost CTF devices to suit a variety of fusion engineering science test missions.« less

New design approaches are proposed for the LHD-type heliotron D–T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

Experiments aiming at 'day long operation at high performance' have been carried out in TRIAM-1M. The record value of the discharge duration was updated to 5 h and 16 min. Steady-state tokamak operation is studied under the localized plasma wall interaction conditions. The distributions of the heat load, the particle recycling flux and impurity source are investigated to understand the co-deposition and wall pumping. The formation and sustainment of an internal transport barrier (ITB) in enhanced current drive mode has been investigated by controlling the lower hybrid driven current profile by changing the phase spectrum. An ITER relevant remote steering antenna for electron cyclotron wave injection was installed and a relativistic Doppler resonance of the oblique propagating extraordinary wave with energetic electrons driven by lower hybrid waves was studied.

An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

Recent progress(1) in plasma science of the Spherical Tokamak (or Spherical Torus, ST)(2) has indicated relatively robust plasma conditions in a broad number of topical area including strong shaping, stability limits, energy confinement, self-driven current, and sustainment. This progress has enabled an extensive update of the plasma science and fusion engineering conditions of a Component Test Facility (CTF)(3), which is potentially a necessary step in the development of practical fusion energy. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes n > 4.4 1013 n/s/cm2, over sizescale > 105 cm2 and depth-scale > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year(4). Such chamber conditions are calculated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~ 3, Ip ~ 9 MA, BT ~ 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. The ST CTF will test the life time of single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A newmore » systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering science test missions.« less

Pellet injection algorithm has been proposed for engineering developments in the force free helical reactor (FFHR) based on the zero-dimensional equations. In general, the plasma parameters, such as the density, the temperature and hence the fusion power, may oscillate by repetitive pellet injections during the ignition access and steady state phase. Pellet injection with a higher frequency cannot solely remove oscillation of the fusion power because it is injected irregularly. Gas puffing with the slow time response using the same signal of the fusion power can compensate these oscillations to some level because pellets are injected cyclically. When two pellet systems work alternatively out of phase together with the slow gas puffing, the fusion power oscillation is smallest. In addition, even when the fuel particle confinement time becomes long due to wall saturation, the fusion power and density can be controlled during the fusion burn phase as long as the helium ash can be removed.

Four types of plasma current ramp-up experiments in full non-inductively lower hybrid current driven (LHCD) plasmas were executed in TRIAM-1M: (1) current start-up by a combination of electron cyclotron resonance heating (ECRH) and LHCD, (2) tail heating by additional LHCD, (3) bulk heating by ECRH and (4) spontaneous ramp-up by a transition to enhanced current drive (ECD) mode. The time evolutions of plasma current during four types of ramp-up phase were adjusted by a simple model with two different time constants, which are a time defined by the total current diffusion time and a time constant for improving the current drive efficiency. In the case of (1) and (4), the latter time constant is significant during the current ramp-up phase. The improvement in the current drive efficiency in the ECD mode is likely to be caused by the increase in the effective refractive index along the magnetic field of the lower hybrid wave.