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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.

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 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.

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.

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.

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.

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.

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.

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.

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.

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.

The present status of experimental and theoretical efforts and future prospects on current start-up and ramp-up methods for ST reactors to reach fusion burning without a center solenoid are discussed. First, we need to form an initial magnetic surface and next to ramp-up the plasma current to a sufficient level for energetic neutral beam injection (NBI). The second heating and current drive by NBI with the ramp-up of the external vertical field may bring the plasma to fusion ignition. The first stage will be realized by refinement of various methods including radio frequency heating and current drive, coaxial helicity injection, plasma merging, and poloidal flux swing. In the second stage, α heating and the bootstrap current dominate as the plasma gets close to burning condition. This is a totally new regime and we need further experimental and theoretical efforts involving novel approaches to ramp-up such a self-regulated plasma successfully.

An overview of steady state tokamak studies in TRIAM-1M (R0 = 0.8 m, a × b = 0.12 m × 0.18 m and B = 8 T) is presented. The current ramp-up scenario without using centre solenoid coils is reinvestigated with respect to controllability of the current ramp-up rate at the medium density region of (1–2) × 1019 m−3. The plasma is initiated by ECH (fundamental o-mode at 170 GHz with 200 kW) at B = 6.7 T, and the ramp-up rate below the technical limit of 150 kA s−1 for ITER can be achieved by keeping the LH power less than 100 kW during the current ramp-up phase. The physics understanding of the enhanced current drive (ECD) mode around the threshold power level has progressed from a viewpoint of transition probability. A transition frequency, ftrans, for the ECD transition is determined as a function of PCD. At ~70 kW no transition occurs for an ftrans value of ~0.017 Hz, meaning almost zero transition probability. With increasing PCD > Pth, ftrans increases up to 10 Hz, and the transition tends to occur with high probability. The record value of the discharge duration is updated to 3 h 10 min in a low and low power (

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved /spl beta//sub T/ of 32%. A variety of MHD phenomena have been observed to limit /spl beta/. NSTX has now begun addressing /spl tau//sub E/ scaling, /spl beta/ limits, and current drive issues. During the NBI heating experiments, a broad T/sub i/ profile with T/sub i/ up to 2 keV, T/sub i/>T/sub e/ and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, /spl tau//sub E/ is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of /spl tau//sub E/ after the transition in some conditions. During HHFW heating, T/sub e/>T/sub i/ and T/sub e/ up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I/sub p/, high q, and high /spl beta//sub p/.

To find a solution of the plasma current rampup problem in a low aspect ratio spherical tokamak (ST) reactor, the effect of the outer vertical field coils on plasma current rampup is studied with noninductively driven current and bootstrap current but without the OH transformer during the fusion power rampup phase. As a lower elongation of [kappa] = 2 does not allow a large poloidal beta, a low density discharge with high heating/current drive power is necessary to increase the noninductive plasma current up to 30 MA, and then the vertical field can ramp the plasma current up to 48 MA just before reaching the steady state fusion burn phase. A higher elongation of [kappa] 3 can reach a higher {beta}{sub p} value, in which case the plasma current of 48 MA can be achieved by the vertical field without powerful heating/current drive. As the level of the vertical field current drive depends on the plasma energy, which is determined by the confinement time, a lower confinement factor can only produce a lower plasma current. The current rampup time can be chosen arbitrarily, shorter or longer, facilitating a flexible ST reactor operation.

Plasma current ramp-up assisted by outer vertical field coils with non-inductively driven current is studied during the ignition access phase in a high aspect ratio tokamak such as ITER-FEAT class tokamak reactor. In a tokamak without the Ohmic transformer, the pl1asma current ramp-up to 8 MA could be achieved by the vertical field with the heating/current drive power of 100 MW. The current ramp-up time can be chosen arbitrarily, as shorter than that expected in the usual, non-inductive current drive operation analyses, and it depends on the fusion power ramp-up time. This makes the reactor operation flexible. Experimental evidence showing the current drive by a vertical field in JT-60U has also been presented with plasma circuit theory.

In tokamak operations, breakdown of plasma is the first step of the plasma build-up. In this paper, we present a combinative investigation of radio frequency (RF)-induced breakdown experiments in QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) and a one-point model of hydrogen ionization. Experimental results with two different frequencies of 2.45 GHz and 8.2 GHz showed that the clear threshold on connection length, L , existed for breakdown with a negative n- index configuration n =(dR/dBv)/(dBv/dR) , where R is the major radius and Bv the is vertical magnetic field. In contrast, breakdown was always obtained with positive n -index when changing L. It indicates that a lifetime of an incubated electron plays a significant role in the plasma breakdown. According to one-point model calculation, the experimental threshold of L is well predicted by the lifetime of the incubated electron estimated by employing the loss term along with L . The model calculation also describes the requirement of the minimum electron temperature Te for RF-induced breakdown to realize an avalanche of electrons in the tokamak magnetic structure.

We compare two cases with and without subtraction of dW/dt in the net heating power after the high confinement mode transition (H-transition) and study the effects of dW/dt on zero-dimensional ignition analyses in various ignition phases in a deuterium–tririum (D–T) tokamak reactor including International Thermonuclear Experimental Reactor (ITER). Subtraction of dW/dt from the net heating power after the H-transition as done in H-mode power threshold analyses results in a wrong control action in feedback control during the confinement factor increment and abnormal behaviors in operation phases such as fusion power ramp-up, steady state, and shutdown phases in a reactor. As subtraction of dW/dt in the L-mode phase is to describe the delay time effect of the heating power to the plasma edge, we conclude that dW/dt should not be subtracted from the net heating power after the H-transition in the zero-dimensional approach. This is because that the H-mode is determined by the net heating power to the plasma edge both before and after the H-transition, but not by the plasma conduction loss power out of the plasma edge.

Plasma current-start up induced by an outer vertical field coil is studied during the ignition access phase in a spherical tokamak reactor. We have illustrated the concept that the plasma current of ~50 MA could be induced by the outer vertical field coil in the proposed spherical tokamak with the help of the small central solenoid flux of ±5 V s and the strong heating power less than 100 MW for the internal inductance of i~0.4-0.8 without the help of bootstrap current and non-inductive current drive power. The required condition to achieve this operation scenario is that the flux produced by the equilibrium vertical field is larger than the inductive flux. Current start-up operation is achieved by adding the small ohmic heating solenoid flux for the flux waveform adjustment because the flux from the outer vertical field coil cannot solely induce the desired plasma current waveform in the case of the preprogramming of the heating power.

The consequence of the failure effect of burn control diagnostic systems, such as neutron diagnostics, bolometers, electron cyclotron emission power loss diagnostics, interferometer, and lost alpha...

Based on the various studies on the ignition characteristics, the feedback control algorithm for ignition burn has been developed for a safe operation and diagnostic integration design of ITER. A set of the diagnostic systems for core burn control are selected and a fail-safe control algorithms has been proposed to prevent the unsafe operation caused by the diagnostic failure. Some feature in the reduced cost of ITER Is also presented.

The main feature of FFHR is force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety. Collaboration works based on the LHD project have made great progress in the reactor studies by focusing on engineering aspects of the high magnetic field and Flibe system design. Encouraging positive results are shown on ignition access, mechanical stress in coils supporting structures, improvement in the blanket system including materials selection and tritium recovery. Critical issues on fundamental safety analysis and maintainability of reactor components are also discussed, and many subjects are pointed out as future works.

Ignition characteristics are studied using zero-dimensional equations and are compared for both the ITER89P and ITER93HP scalings with the H-mode power threshold in the International Thermonuclear Experimental Reactor-intermediate design (ITER-ID). With the help of the operation path method on the -T plane, the ignition behaviour for both scalings is compared. The comparison study illustrates that the confinement factor of 2.6-2.9 over the ITER89P scaling is necessary for obtaining the equivalent performance to the ITER93HP scaling with 85% degradation due to an ELMy discharge. If the confinement degradation effect of due to the radiative divertor operation is taken into account, the confinement factor should be increased by up to . This result implies that the `advanced' tokamak operation regime may be required even for MA nominal operation from the outset. Modifications of the plasma shape, the central solenoid and divertor would be necessary to increase the confinement time for reducing the ignition risk. The ignition behaviour for both scalings are entirely different, where the ITER93HP scaling is more favourable for the safer ignited operation. Thus, in this sense, ITER89P scaling is highly recommended for designing the ITER.

The temporal evolutions of the operating point during the ignition access and ignited operation phases are analysed on the basis of zero dimensional (O-D) equations in order to clarify the requirements for safe control of ignited operation and for the development of diagnostic systems in ITER. A stable and safe method of reaching the ignited operating point is identified as the `higher temperature access' method, being compatible with the H mode power threshold constraints. It is found that the ignition boundary can be experimentally determined by a `thermonuclear oscillation' of the operating point without knowing the power balance equation. On the other hand, the ignition boundary determined by the power balance equation has a larger error bar depending on the accuracy of the diagnostic system. The plasma waveform response to sudden changes in the various plasma parameters during ignited operation is also calculated, and fusion power regulation is demonstrated by feedback control of the fuelling and auxiliary heating power

Complete one cycle and 1.5 cycle ac operations are performed in the STOR-M tokamak with the plasma current of ∼20 kA using newly developed feedback control and Ohmic heating circuits. Bias voltage adjustment is installed in the plasma position circuit to optimize the plasma position in the second negative plasma current phase for multicycle ac operation. The key to successful, reproducible multicycle ac tokamak operations on STOR-M is to control both the total vertical field by the feedback control system and the plasma position by application of the bias voltage.

The ignition access and ignited operation regime are studied using a POPCON analysis with the H-mode power threshold scalings for a D - T experimental tokamak reactor, such as the International Thermonuclear Experimental Reactor (ITER). The H-mode power threshold scaling provides two H-mode regimes in the low-density and high-density regions. The L-mode regime exists between them. As the two H-mode regimes are connected at a plasma temperature around 30 keV, the operation path must make a detour around this connecting area to reach ignition within a reasonable heating power. Although the heating power to reach ignition depends on the coefficient in the H-mode power threshold scaling and plasma parameters, the heating power of 50 MW, as originally proposed by Rebut et al, is marginal and not sufficient to reach ignition for the ITER tokamak even if the L - H transition can be first achieved in the initially low-density regime of less than . Auxiliary heating power of 100 MW may satisfy the various requirements for D - T ignition experiments in the ITER tokamak. The thermal equilibrium operating point with fusion power of 1.5 GW in the ignition regime can exist only when the H-mode power threshold has a hysteresis for the ASDEX Upgrade threshold scaling.

Feasibility of D-3He fusion reactor has been investigated by using 0-dimensional power and particle balance equations and the IPB98(y,2) scaling. A self burning spherical tokamak (ST) D-3He reactor requires the high Ti mode and low impurity contamination. The beta value, the confinement improvement factor (HH98y2) and the superconducting magnet should be improved from the present level.

Ignition characteristics in deuterium-tritium helical (stellarator) reactors of various sizes are studied with the operation path method on the plane and the POPCON method. Based on empirical large...

Preliminary key issues for a synchrotron radiation-enhanced compact fusion advanced Brayton (CFAB) cycle fusion reactor similar to the CFAR (compact fusion advanced Rankine) cycle reactor are presented. These include plasma operation windows as a function of the first wall reflectivity and related issues, to estimate an allowance for deterioration of the first wall reflectivity due to dpa effects. It was found theoretically that first wall reflectivities down to 0.8 are still adequate for operation at an energy confinement scaling of 3 times Kaye-Goldston. Measurements of the graphite first wall reflectivities at Nagoya University indicate excellent reflectivities in excess of 90% for CC-312, PCC-2S, and PD-330S in the submillimeter regime, even at high temperatures in excess of 1000 K. Some engineering issues inherent to the CFAB cycle are also discussed briefly in comparison with the CFAR cycle which uses hazardous limited-resource materials but is capable of using mercury as coolant for high heat removal. The CFAB cycle using helium coolant is found to achieve higher net plant conversion efficiencies in excess 60% using a non-equilibrium magnetohydrodynamic disk generator in the moderate pressure range, even at the cost of a relatively large pumping power, and at the penalty of high temperature materials, although excellent heat removal characteristics in the moderate pressure range need to be guaranteed in the future.

Formation of a high performance plasma with βp = 3.6, βN = 1.6, HH98y2 = 1.6 without the use of the Ohmic heating solenoid was demonstrated for the first time. Plasma start-up and plasma current ramp-up were accomplished by vertical field and shaping coils, combined with heating and current drive by EC, LH and NBI. The q profile is deeply reversed with qmin = 5.6 at r/a = 0.7, and the bootstrap current fraction was at least 90%. This result opens up the possibility of eliminating the Ohmic heating solenoid, which has a great impact on improving tokamak and ST reactor designs.

The 16-element phased-array antenna has been developed for the EBWH/CD experiments. The 2-D k-spectra could be controlled. The heat load and stress were analyzed to be moderate in the 200kW CW operation. The symmetrically 2-D narrow-field distributions were excited by the developed antenna, indicating its good applicability to the EBWH/CD experiments.

The current rampdown phase followed by non-disruptive current termination is studied for future alternating current (AC) operations in the STOR-M tokamak. The plasma current is smoothly reduced to zero without disruptions with the rampdown rate of around 1.5 kA ms-1. The plasma density in the current rampdown phase is controlled by gas puffing. The plasma current decreases, maintaining the relationship 1/qa approximately 0.01 nR/Bt (10-18 m2T). At current termination, Ip=0, a residual plasma density (0.58-2.9)*1018 m-3 is observed with a corresponding Murakami parameter nR/Bt=(0.41-2.1)*1018 m-2 T-1. After current termination, the low-density plasma remains for 5 to 15 ms, which would allow AC operation in the STOR-M tokamak if the radiation barrier does not affect the subsequent current evolution.

It is shown that a tokamak with a major radius larger than [approximately]6 m and a toroidal field of 10 T can reach ignition by ohmic heating alone at a relatively low peak density [n(o)[approximately]1[times]10[sup 20] m[sup [minus]3]] even with confinement degradation due to alpha-particle heating, provided a confinement enhancement factor of [gamma][sub H] = 2 over the Goldston scaling is assumed. The critical toroidal field and plasma current required for ohmic ignition have been estimated for various sizes of tokamaks with major radii R = 2 m (compact), 6 m [International Thermonuclear Experimental Reactor (ITER) class], and 10 m (large tokamaks). If a broad current profile can be achieved transiently, the critical toroidal field and plasma current can be further reduced by the enhancement in the ohmic heating power. 37 refs., 7 figs.

In this paper, the method of the operation path on the generalized ignition contour map ({bar P}{sub ht} {tau}{sup 2}{sub E}, n{tau}{sub E}, T) is used to treat the ignition accessibility of a deuterium-tritium tokamak reactor for different types of confinement scaling including offset linear scaling. It is shown that the operation paths and ignition boundary on the {bar P}{sub ht}{tau}{sup 2}{sub E}-T plane as well as the n{tau}{sub E}-T plane provide a useful ignition criterion for various types of scaling. The International Thermonuclear Experiment Reactor (ITER) tokamaks with low and high toroidal fields are used as examples.

In this paper, ignition characteristics in deuterium-tritium(D-T) and D-{sup 3}He tokamak reactors with spin-polarized fuels are presented by using the ignition access condition based on the generalized saddle point in the representation of ({bar P}{sub ht}{tau}{sup 2}{sub E}, n{tau}{sub E}, T). Enhancement of the D-T fusion cross section due to parallel spin polarization with respect to the magnetic field can reduce the confinement enhancement factor required for reaching ignition by {approximately} 20% if fusion particle loss is not induced by the anisotropic fusion particle distribution. Spin polarization is this effective when a D-T reactor is marginal for ignition. In D-{sup 3}He fusion, it is more advantageous to use spin-polarized fuel in the heating phase than in the case of D-T fusion. The auxiliary heating power to reach ignition, which is rather large for D-{sup 3}He fusion, can be reduced by a factor of 2 to 3 compared with the unpolarized case. The deuterium-deuterium fusion suppression effect, it is exists, does not alter the ignition condition much. Various problems related to the spin polarization scheme are also discussed.

The phased-array antenna has been designed for steady-state electron Bernstein wave heating and current drive at 8.2 GHz in the QUEST tokamak. The prototype antenna had rather large side-lobe components. In the new designed antenna the side-lobe components were reduced by modification of the antenna structure. For the steady-state operation, forced water-cooling and thermal stress were analyzed for the antenna component using the HFSS/ePhysics/ANSYS codes.

Recent progress of a D-T tokamak fusion reactor with advanced blanket concept using CFAR (Compact Fusion Advanced Rankine) cycle is presented. High enthalpy extraction in excess of 50% was found achievable by a nonequilibrium disk-type MHD generator for gas stagnation temperatures of 3000 K superheated by synchrotron radiation, leading to a 40% overall plant efficiency with recuperative heat cycle and advanced thermoelectric converters. The stored magnetic energy of a 10 T MHD three-pairs Helmholtz-type magnet was estimated to be about 600 MJ per MHD generator. A mixture of mercury and lithium liquid flow was examined and found applicable to reduce the mercury inventory. The tritium permeation loss inherent in the relatively high temperature CFAR cycle, neutronics relating to tritium breeding and streaming through the waveguide and the MHD generator duct are discussed. Results of a 2000 h high temperature (600–700°C) mercury corrosion test of Type 304 stainless steel, a TZM alloy and four kinds of ceramic rods in refluxing mercury with 0.004 mol fraction of potassium admixture have shown no noticeable corrosion except for Type 304 stainless steel.

In this paper an alternating current (ac) tokamak reactor with ohmic ignition and long pulses due to bootstrap current is proposed as a simple and quasi-continuous fusion power plant. An ohmic plasma current of 23 MA with a high toroidal field of {approximately} 10 T in the alternating Current Tokamak Reactor-Upgrade (ACTR-U) (10-m major radius and 2-m minor radius) provides the ohmic ignition. After entering the ignition regime, the plasma current is reduced by one-half to enhance the bootstrap current with a high-beta poloidal field ({beta}{sub p} {approximately} 2) to prolong the pulse length. When the ohmic transformer reaches the maximum flux, the plasma current is ramped down and reversed; ac operation follows. The authors thus demonstrate that an ohmic transformer alone is in principle sufficient for a quasi-continuous deuterium-tritium fusion reactor.