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The mission of the National Centralized Tokamak (NCT) is to establish high-beta steady-state plasma operation for DEMO and to contribute to ITER. Plasma performance with high-beta conditions is discussed. It is found that near break-even class plasma ( Q DT eq ≥ 0.7 ) with high beta ( β N ≥ 3.5) is achievable with 25 MW of NB power at I P = 4.0 MA and an aspect ratio of 2.6. High beta with steady-state full current drive is possible with 13 MW of NB power at I P = 1.5–1.7 MA. In the engineering design activity of NCT, the strength of the support structure and strength of material for the conductor is confirmed for the superconducting coils, and the successful progress of R&D of boron-loaded resin for neutron shielding is discussed.

The present fusion program of Japan is based on the ''Third Phase Basic Program of Fusion Research and Development,'' determined by the Japan Atomic Energy Commission in 1992. This paper describes the Third Phase Basic Program and the overview of the relevant present research status.

Direct measurements of localized electric fields have been made by the laser induced fluorescence (LIF) method using the Stark effect in the central cathode core region of an inertial electrostatic confinement fusion (IECF) neutron (proton) source. These are expected to have various applications, such as luggage security inspection, non-destructive testing, land mine detection and positron emitter production for cancer detection, currently producing continuously about 107 n/s D-D neutrons. Since 1967, when the first fusion reaction was successfully proved to have taken place in a very compact IECF device, potential well formation due to the space charge associated with spherically converging ion beams has been a central key issue remaining to be clarified in beam-beam collision fusion, which is the major mechanism of the IECF neutron source. Many experiments, although indirect, have been done so far to clarify the nature of the potential well, but none of them has produced definitive evidence. The results found by the present LIF method show a double well potential profile with a slight dip for ion beams with relatively larger angular momenta, whereas for ions with smaller angular momenta, a much steeper potential peak develops.

The effects of beam convergence on the fusion reaction rate in the cylindrical inertial electrostatic confinement fusion device are investigated using a two-dimensional simulation code and experiments. It is found from the simulation that the fusion reaction rate increases significantly with an increase of beam convergence; therefore, there is an increase of ion densities at the center region. In the experiments designed to confirm these results using different electrode shapes, the effects of the anode shape are clearly observed.

Strongly localized electric fields were measured in the central cathode helium plasma core region of an inertial electrostatic confinement fusion device by using laser-induced fluorescence (LIF) by the degree of polarization and by the longitudinal alignment methods. Both results show double well potential formation with a slight concave at the center in excellent agreement. The decay time of the excited states is found to indicate least effects by the collisions to ensure the LIF method.

A large-bootstrap-current tokamak plasma driven by the ohmic seed current has been studied for inductively-operated ultra-long-pulse tokamak fusion reactors. The safety factor profile with the negative magnetic shear has been demonstrated only by the combination of the ohmic and bootstrap currents, by adjusting the density profile, and the bootstrap current fraction more than 80% is realized with the help of the increased q(0) value. 7 refs., 4 figs., 1 tab.

The Large Helical Device (LHD) is a big non-tokamak type superconducting toroidal fusion facility. In both areas of physics and technology, the LHD project is expected to contribute to the exploration of necessary understandings of the currentless steady-state plasmas, which have high performance of temperture, density, confinement time and beta value extrapolatable to the reactor regime. This paper describes the necessary issues of major operation scenario especially concerned with the divertor functions, which realize the steady-state experiment in LHD. 16 refs., 8 figs., 2 tabs.

As an intense 14-MeV neutron source, a steady-state subignited tokamak plasma is proposed, where a 60-MW neutral beam is injected to sustain a subignited plasma and to drive a plasma current for steady-state operation. Plasma and device parameters are self-consistently designed, taking into account physical (confinement characteristics, beta limit, current drive efficiency, and so on) and engineering (maximum magnetic field strength, blanket/shield thickness, and others) constraints. The result of a comparison between plasmas with A = 2.8 and A = 4 indicates that a large aspect-ratio device is preferable as a neutron source. A surface-averaged 14-MeV neutron flux of {approximately} 0.6 MW/m{sup 2} is achievable with R = 4 to 5 m, A = 4, and B{sub max}= 10 T and is not so sensitive to the major radius. When the maximum magnetic field strength of toroidal field coils is raised to 13 T, a neutron flux more than 1 MW/m{sup 2} is available with a device with R = 4 m. If the plasma performance is advanced and plasmas with an L-mode enhancement factor f{sub L} of {approximately}3 and a Troyon coefficient in beta limit g of {approximately}5 are attainable, a neutron flux of {approximately} 1.6 MW/m{sup 2} ismore » achievable even with a device with R=4 m and B{sub max} = 10 T. These devices seem to be very attractive not only as a neutron source but also as a supplementary device of an ignition-oriented International Thermonuclear Experimental Reactor (ITER) device. 16 refs., 7 figs., 1 tab.« less

The plasma and machine parameters of a pulsed tokamak reactor with a day-long operation period have been studied, where engineering constraints such as maximum toroidal field strength are preserved at International Thermonuclear Experimental Reactor (ITER) levels so as to realize a fusion reactor with only a short-range extension of currently available technology. To provide the magnetic flux necessary to sustain a plasma current inductively for 1 day or longer, plasma with a major radius of R > 9.5 m are necessary, and a plasma with an aspect ratio as high as A>5 should be employed. Typical parameters are as follows: major radius R = 10 m, minor radius a = 1.85 m, plasma elongation [kappa] = 1.8, plasma current I[sub p] = 12.2 MA, toroidal field on axis B[sub t] = 7.56 T, and safety factor at the plasma surface q[psi] = 3. A plasma volume V [approximately] 1200 m[sup 3] is comparable with that of ITER, even though the major radius of a day-long operation reactor is relatively large. A very small amount of heating power ([approximately]15 MW) with a heating time of only a few tens of seconds is sufficient to achieve the ignition condition. This is wellmore » within the capacity of auxiliary heating systems currently used in large tokamak devices. A confinement improvement factor (from L mode) of [line integral][sub L] > 1.7 is required to design a reactor with a reasonable machine size and a day-long pulse duration. The operation temperature is chosen to be [r angle]T[l angle] = 20 keV with a toroidal beta [beta][sub t] = 2.6% (Troyon factor g = 3), which gives a fusion power P[sub fus] = 2.5 GW even for an alpha-particle dilution n[sub [alpha]]/n[sub e] of 10%. The bootstrap current fraction is 50% or more of the total current, and current profile needed for the beta limit could be achieved with a combination of ohmic current in the plasma center region and bootstrap current in the outer region. 23 refs., 8 figs., 2 tabs.« less

Ohmic ignition condition on axis of the DT tokamak plasma heated by minor radius and major radius adiabatic compression is studied assuming parabolic profiles for plasma parameters, elliptic plasma cross section, and Neo-Alcator confinement scaling.It is noticeable that magnetic compression reduces the necessary total plasma current for Ohmic ignition device.Typically in compact ignition tokamak of the minor radius of 0.47m, major radius of 1.5m and on-axis toroidal field of 20T, the plasma current of 6.8MA is sufficient for compression plasma, while that of 11.7MA is for no compression plasma. Another example with larger major radius is also described.In such a device the large flux swing of Ohmic transformer is available for long burn.Application of magnetic compression saves the flux swing and thereby extends the burn time.

The resistance anomaly in ultra-low q (surface safety factor q(a) < 1) discharges has been studied experimentally. The effective resistance (loop voltage/plasma current) is much higher than the kinematic resistance calculated with the electron temperature. The effective resistance is inversely proportional to the applied toroidal magnetic field. This scaling applies universally to devices with different dimensions.

Real time measurements of the electric fields were made in the central cathode core region of an inertial-electrostatic confinement fusion (IECF) device by the laser-induced fluorescence method by use of Stark effects. Preliminary results show double well potential formation with a slight concave at the center and also the existence of energetic electrons in the localized region near potential peaks.

IECF experiments with auxiliary discharges in outer region were made. Operation pressure of IECF is successfully reduced to about 0.5 Pa when auxiliary discharges are turned on. Neutron measurement shows 2/spl times/10/sup 6//sec fusion occurred at 45 kV, 15 mA discharge. Proof of principle experiment of the potential well measurement using Stark effect shows the method works but several improvements are required.

An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.

Since a neutron has a magnetic moment, it interacts with a magnetic field. A possible configuration to confine ultracold neutrons (UCN) is a combination of a toroidal multipole field and a toroidal field; the latter prevents spin flip at the separatrix of the multipole field. It results in effective cooling to apply an appreciable ripple to the toroidal field which propagates at the resonant phase velocity. The cooling rate which is the power removed by a propagating toroidal magnetic wave was obtained through the quasilinear theory of the transit-time dumping for a plasma, and compared to the particle simulations.

A three-dimensional particle simulation scheme has been developed and applied to the study of relativistic charged particle beams. To simulate the nonlinear behaviors of beam electrons with self-induced space-charge effects, it simultaneously solves the canonical equations of motion and the integrated Maxwell equations of radiated waves for each particle. Verifications of this algorithm show that calculations for the retarded structure of radiated waves dominate the total throughput, and the near-field approximation to the calculations of space-charge potentials improves its numerical efficiency. The code has been applied to a simulation of the Raman regime free-electron laser (FEL) to demonstrate that this scheme simulates the parametric amplification of FEL properly and is suitable for relativistic treatments of such wave-particle resonances.

A mathematical method to analyze the decay of a dissipative dynamical system is developed, which is to study the dissipative structure of the corresponding system. The method is applied to the study of stable equilibria in magnetohydrodynamic (MHD) systems. Special sets of MHD equilibria are characterized as attractors of MHD systems, and they are shown to be classified into four classes, which cover wide range of experimentally observed MHD equilibria.

The characteristics of plasmas with the safety factor q between that of tokamaks and that of reversed field pinches were studied on the OHTE device, with toroidal fields of up to 3.9 kG and plasma currents of up to 140 kA. Plasmas in this regime (0 < q < 1) are called ultra-low q plasmas. The discharges are globally stable and last much longer than the resistive decay time of the current. A unique feature of most discharges is a series of plateaus in the current (or equivalently q) as the current decreases in a 'staircase' fashion. The plateau q is related to the major rational ratios (1/4, 1/3, 1/2, 1, etc.), suggesting plasma self-organization. Profile measurements show that the q-profile is somewhat different from those of the Taylor relaxed states. The soft X-ray temperature measurement indicates that the plasma has burned through the radiation barrier.

Bursts of impurity radiation in the visible spectral region are observed in the sustainment phase of the STP-3M reversed-field-pinch plasma. The phenomena occur simultaneously with enhancement of the toroidal magnetic flux relevant to the dynamo action and sudden increase of the ion density in the peripheral region of the plasma. Compared with calculated results using a time dependent 0-D corona model, the measured bursts of impurity radiation interpret the enhancement of impurity influx due to the plasma/wall interactions accompanied by the dynamo action.

MHD relaxation phenomena have been studied for a toroidal plasma in a safety factor q

Structure mathematique de la quasi-statique de plasmas bi-dimensionnels porteurs de courant

Runaway electrons have been investigated in a tokamak with an electric fie1d up to 0.5 Vcm-1 which is smaller than the Dreicer field. Dependences of the X-ray yield on the filling pressure P0 and the electric field E are investigated experimentally. The runaway electron flux density is calculated as a function of time for various values of P0 and E. Agreement between experimental results and calculations is fairly good. Both the X-ray intensity and plasma current are measured as a function of the vertical field Bv. The optimum vertical field for the maximum runaway current can be explained by considering a single particle drift orbit. In the case of a higher vertical field a considerable part (more than 10%) of the plasma current is carried by runaways.

The energy loss by radiation is estimated in REPUTE-1 by bolometric measurements. Burning out of the radiation barrier is observed in both ultra-low-q (ULQ) and reversed field pinch (RFP) discharges. The conditions for burning out of the radiation barrier are studied for various operation parameters. A reduction of the radiation loss is observed after wall conditioning by carbonization.

Recent theoretical contributions to magnetohydrodynamics (MHD) have been reviewed from a view-point of mathematical sciences. Analytical structures of the convective nonlinearity in the incompressible dissipative MHD have been discussed stressing on their relevance to the fundamental theory of dynamical systems.

Three‐dimensional codes for calculations of a stray magnetic field of iron core of a toroidal plasma confinement system have been developed. The calculated results agree well with the measured ones. It is found that up‐and‐down asymmetry of the iron‐core gap deforms the stray field asymmetrically when the primary coil current is not equal to the plasma current and that anisotropy of permeability of the iron core does not have a large influence on the field.

Internal modes with m = 1 and m ≥ 2, which localize around a pitch minimum, have been studied for ultra-low-q plasmas with 1/2 ≤ qa < 1 in REPUTE-1, where m is the poloidal mode number and qa is the safety factor at the plasma surface. One-dimensional stability analysis has shown that, for finite beta, m ≥ 2 modes have a relatively large growth rate compared with those of the m = 1 modes. The internal structure of the magnetic fluctuations with m = 1 and m ≥ 2 is found to be consistent with the numerically calculated eigenfunctions.

This paper studies a conceptual design of a compact high-energy neutron source which might be considered for applications in the field of fusion-material researches. A possibility of a compact neutron source has been found by matching advantages of the ULQ equilibrium and the beam-driven (TCT) concept. The ULQ configuration, which is in the safety-factor

The loop voltage of REPUTE-1 RFP is decreased by reducing the toroidal field ripple. The low frequency part of the toroidal field fluctuations, which is mainly composed of the m = 1 mode and has good coherence in space and time, is found to be reduced, especially outside the reversal surface, when a trimming field is applied. High frequency fluctuations, which have little coherence, are not influenced effectively.

Eddy current analysis for REPUTE-1, which is a Reversed Field Pinch (RFP) experimental device at the University of Tokyo, is presented. The error field problem is one of the most serious ones for plasma confinement in RFP. When the plasma current starts up, a large amount of eddy current flows in the plasma vessel. The plasma vessel consists of an inner vacuum vessel and an outer toroidal shell. The plasma vessel has many holes, for example, 108 holes in REPUTE-1, for vacuum pumping and plasma diagnosis. The flow pattern of the eddy current is badly distorted due to the existence of these holes. This distortion produces large ripples in the magnetic field generated by the eddy current in both the toroidal and poloidal directions; this is termed an error field. So far, the plasma confinement time has been measured as rather short compared to a design value in RFP experimental devices elsewhere because of the error field. In this analysis, we used the software tool named INCANET (IHI Induced Current Analysis Code by the Network Mesh Method). The toroidal and poloidal distributions of the error fields were numerically obtained by the three-dimensional eddy current analysis. When the total eddy current in the vacuum vessel is 30 kA, which corresponds to the typical RFP discharge of 200 kA plasma current, the radial component of the error field is 0.014–0.015 T at the edges of port holes, which is around 8% of the poloidal field generated by the plasma current itself.

We study an analytical structure of a quasistatic model of magnetically confined plasmas. Applying the fixed point theorem, we construct global‐in‐time weak solutions.

The characteristics of a toroidal discharge with qa ≈ 1/2 is studied on TORIUT-6. Ultra low q equilibria that are stable against the global kink mode are realized through MHD relaxation and are shown to be characterized by dq/dr < 0 and a hollow current profile. The stable state lasts for 600 μs, which is of the order of the classical diffusion time. These experimental observations are consistent with non-linear MHD simulation results.

Temperature gains due to three types of adiabatic toroidal compression are compared on the assumption of the fixed chamber dimension and the maximum toroidal field on the chamber axis. A combined adiabatic heating scheme may be more advantageous in obtaining a high-temperature plasma than minor or major radius compressions depending on the plasma loss mechanism. A combined adiabatic compression scheme is more favourable than a major radius compression scheme when the final plasma dimensions are the same, for all plasma loss scalings considered here.

Toroidally asymmetric perturbations of toroidal flux in a reversed field pinch (RFP) plasma have been observed with a set of 18 flux loops in ramping current discharges on the REPUTE-1 RFP. The toroidal flux perturbation is induced toroidally localized and spreads as it propagates in the electron-diamagnetic drift direction. The dominant toroidal mode number of the flux perturbation is usually 1, but becomes 2 when the toroidal flux is strongly generated, and the second peak merges with the first one as it propagates in the toroidal direction. Thus, a locally generated toroidal flux becomes toroidally symmetric.

The ideal MHD stability of non-monotonic q-configurations has been studied for an ultra-low-q (ULQ) regime (0 < q < 1) by both analytical and computational methods. An ULQ configuration is characterized by a safety factor q-profile with dq/dr < 0 in a part of the plasma region and the existence of a pitch minimum. An ULQ equilibrium for single resonant surface modes with m = 1 is stable because of the wall stabilization effect, if the equilibrium has enough shear. Modes with m ≥ 2 are localized around the resonant surface and are stable. Double resonant surface modes are unstable, but the growth rates of these modes are smaller than those of internal kink modes. The stability of m ≥ 2 modes is strongly dependent on the depth of pitch minimum.

This paper studies the energy-balance characteristics of an ultra low-q (ULQ) confinement concept, and shows the possibility of the Ohmic-Heating ignition. The ULQ equilibrium considered here has the safety factor q in the intermediate range 0< q< 1 lying between tokamaks (1

Etude du comportement asymptotique de plasmas incompressibles, introduisant une approche deterministe a la dynamique d'un systeme dissipatif avec non-linearite convective. Le systeme a des etats quasi stationnaires non triviaux qui correspondent aux equilibres de tokamak et de striction a champ inverse

A current-drive scheme based on combining neutral-beam injection and ICRF wave heating is investigated. The neutral-beam injection mainly supplies toroidal momentum and is sustained by ICRF wave heating. The current is generated with sufficient efficiency and can sustain the tokamak plasma in a steady state.

高速中性粒子入射加熱装置においては, イオン源から引き出されたビームが, 中性化されたのち標的プラズマに打ち込まれる。このとき孔軸偏位した多孔引出電極を用い, 全ビームレットを標的ポートに集めるという方法をとれば, ビームラインにおける粒子輸送効率が上がる。孔軸偏位した電極の設計には, 三次元のイオンビーム引出シミュレーションコードが必要となる。この技術報告では, 三次元コードの定式化, アルゴリズム, 計算結果, 及びこのコードの他分野への応用の可能性について述べている。

Magnetic field measurements in a reversed field pinch plasma have been made with an insertable magnetic probe in REPUTE-1. The relaxation process of the magnetic field configuration in a high θ (θ is the pinch parameter) regime has been studied. The safety factor on axis, q0, decreases as θ increases, and there exists a lower limit to q0(~0.10). Close to this limit, q0 increases when the magnetic field configuration relaxes. These observations agree well with numerical results of non-linear resistive MHD theory.

Soft X-ray measurements have been made with surface barrier diode (SBD) arrays in the REPUTE-1 RFP device (R/a = 82/22 cm). Arrays of seven and eleven SBDs are arranged so that they view a minor cross-section of the plasma from the vertical and horizontal diagnostic ports, respectively. Applying the tomography technique to the soft X-ray data, a two-dimensional emissivity profile can be reconstructed. In the sustainment phase of the RFP discharge, the soft X-ray emissivity profile exhibits a periodic peaking and a subsequent large crash across the minor cross-section. The fluctuation period is in the range of 60–150 μs, whereas the duration of the large crash is always about 30 μs (corresponding to ~ 100 × τA, where τA is the Alfven transit time). The soft X-ray emission is due to two modes: one is a dominantly m = 0 mode, which corresponds to the axisymmetric increase or decrease of the soft X-ray emissivity profile, and the other is an m = 1 mode. The toroidal correlation length of the m = 0 mode ranges from 20° to 80°, while the m = 1 mode has a high toroidal mode number. In the crash process, an m = 1 distortion of the emissivity profile, which indicates reconnection of the flux surfaces, can be observed. The magnetic fluctuations observed simultaneously are well correlated with the soft X-ray fluctuations. It is found that the soft X-ray fluctuations reflect the relaxation mechanism of the RFP plasma. This phenomenon is considered to be driven by global MHD activity.

Elastic and inelastic scattering of 28.5 MeV alpha-particles have been investigated. Blair's phase rule is fairly well satisfied in forward angles (

The principle of Project MACETIE and the preliminary experiment are described. The basis of our plan is the thought that the goal of controlled thermonuclear fusion may be accessible by the multistage alternating processes of adiabatic compression and irreversible expansion of plasma. In the latter process, the plasma compressed by a strong magnetic field is expanded into a region of weaker magnetic field. A static consideration of this expansion suggests that plasma temperature will not change appreciably in this process, and a second adiabatic compression of plasma in the weak field may further increase its temperature. Preliminary experiments are being carried out to examine the foregoing principle of irreversible expansion.

The excitation of unnatural-parity states (2/sup -/, 3/sup +/, etc.) was investigated at E/sub alpha / = 28.5 Mev on /sup 16/O, /sup 20/Ne, /sup 24/Mg, and /sup 28/Si at forward angles smaller than 90 deg (lab). Angular distributions of the inelastically scattered alpha particles are shown. Possible reaction mechanisms are discussed. (R.E.U.)

The angular distributions of the reactions Ni 58 (p, α) Co 55 , Ni 60 (p, α) Co 57 and Fe 56 (p, α) Mn 53 leading to the ground and several excited states of the residual nuclei have been measured at 12.6–14.25 MeV, at 10.02–14.43 MeV and at 11.0 MeV, respectively, making use of solid state detectors. Levels have been found in Co 55 at 1.8 ± 0.2, 2.2 ± 0.2 and 2.45 ± 0.2 MeV and in Co 57 at 1.77 ± 0.02, 2.08 ± 0.03 and 2.21 ± 0.03 MeV. It was found that at about 14 MeV the forward peaks are predominant in the angular distributions for the reactions Ni 58 (p, α) Co 55 and Ni 60 (p, α) Co 57 leading to the ground states of the residual nuclei. The forward peaking disappears gradually with decreasing proton energy. The angular distributions of the reactions Ni 58 (p, α) Co 55 and Ni 60 (p, α) Co 57 leading to the several excited states of the residual nuclei have approximately 90° symmetry. The angular distributions of the reaction Ni 58 (p, α) Co 55 leading to the ground state of Co 55 have been measured at 28 energies between 8.96 and 10.32 MeV. Sharp variation of the shape of the angular distribution with the bombarding energy was obtained. The results of the measurements have been analysed with the theory of the compound nucleus developed by Ericson.

We show some experimental results for solidification of CO2 and the dropping in vacuum. It is a preliminary experiment for solid D2 dropping in vacuum, in order to produce high density and high temperature plasma with giant pulse laser.

The expansion process of a collision-dominated theta-pinch plasma in a homogeneous magnetic channel and the thermalization process of the kinetic energy of this plasma flow by a magnetic mirror field are studied on the BSG-II device. Previously it was reported that shock phenomena due to a mirror field were observed by several methods of measurement. In this paper four experimental results are presented: (1) the expansion of plasma along the homogeneous magnetic channel is adiabatic; it is well described by a model for one-dimensional isentropic flow of collision-dominated plasma; (2) the flow is supersonic and the occurrence of a shock in front of the mirror field is in accordance with a criterion concerning the supersonic-sonic transition of plasma flow; (3) the rate of plasma loss through the mirror field is reduced by the occurrence of the shock; (4) the values of the plasma parameters measured before and behind the shock front satisfy the Rankine-Hugoniot relations, though plasma pressure, flow velocity and Mach number are time-dependent.

The purpose of the experiment using the BSG-I device described here is to study the expansion of a plasma, produced in a strong magnetic field by a theta pinch, into a weak magnetic channel and the thermalization of the kinetic energy of this plasma flow by a magnetic mirror field. The plasma, to which the fluid model is applicable, treated here is collision dominated and low-β. Shock phenomena due to the mirror field at the far end of the device are observed by several kinds of measurements and it was concluded that some part of the kinetic energy of the plasma flow was changed to thermal energy by the shock phenomenon. Plasma parameters measured in front of and behind the shock front were consistent with the Rankine-Hugoniot relations. Thus the ion temperature rises after the shock and approaches that of the initial plasma as produced in the theta pinch region before the expansion.