Toshihide Tsunematsu

Mechanism of vertical displacement events in JT-60U disruptive discharges

Enhanced vertical displacement events (VDEs), which are frequently observed in JT-60U disruptive discharges, are investigated using the Tokamak Simulation Code (TSC). The rapid plasma current quench can accelerate the vertical displacement, owing to both the up/down asymmetry of the eddy current distribution arising from the asymmetric geometry of the JT-60U vacuum vessel and the degradation of magnetic field decay index n, leading to high growth rates of positional instability. For a slightly elongated configuration (n=-0.9), the asymmetry of attractive forces on the toroidal plasma plays a dominant role in the VDE mechanism. For a more elongated configuration (n=-1.7), the degradation of field decay index n plays an important role on VDEs, in addition to the effect of asymmetric attractive forces. It is shown that the VDE characteristics of a highly elongated configuration with a rapid plasma current quench can be dominated by the field decay index degradation. It is also pointed out that both the softening of current quenches as was experimentally developed in the JT-60U tokamak, and the optimization of the allowable elongation of the plasma cross-section are critical issues in the development of a general control strategy of discharge termination

Study of the density limit with pellet fuelling in JT-60

The density limit for a series of gas fuelled and pellet fuelled limiter discharges in JT-60 has been studied. With pellet injection into high current/low q (qcyl = 2.1–2.4) discharges, the Murakami factor reaches (10–13) × 1019 m−2·T−1. The values are factors of 1.5–2.0 higher than those for gas fuelled discharges. For pellet fuelled discharges the central density is high, whereas in the outer region (a/2 < r) the electron density is limited to the same level as that for gas fuelled discharges. The density limit is confirmed to be an edge density limit; this can be explained by the power balance in the outer region of the plasma, for which the plasma purity (Zeff), the heating power (Pabs) and the electron temperature are the key parameters. The onset of a disruptive event can occur when Prad(total) ~ 12–80% of Pabs(total), and the disruptive limit of the density can be explained by Pabs and ne2 (r = 50 cm) Zeff (r = 50 cm). The thermal stability in the edge region is weaker for lower (<0.4 keV) edge electron temperature.

Second stability region against the internal kink mode in a tokamak

Using the ERATO code, the stability of the internal kink mode is studied for a wide range of βp values for tokamak equilibria with a circular cross-section. The internal kink mode becomes unstable when βp exceeds a critical value and becomes stable again when βp is sufficiently high.

Ideal MHD stability of pellet fuelled plasmas in JT-60

Stability against ideal infinite-n ballooning and low-n internal kink modes is analysed for pellet fuelled plasmas in JT-60, which have strongly peaked pressure profiles within the q = 1 resonance surface. The observed pressure gradient in pellet fuelled plasmas locally reaches the marginal value for the infinite-n ballooning mode limit just inside the q = 1 surface. After reaching the marginal value on the magnetic flux surface, the marginally stable region grows inside the q = 1 surface. However, the poloidal beta within the q = 1 surface, βp1, does not reach the limit imposed by the ideal ballooning mode. The results of the analysis of the low-n kink mode using experimental observations show that the maximum βp1 attained is consistent with the beta limit imposed by the n = 1 internal kink mode.

Approximation of eddy currents in three dimensional structures by toroidally symmetric models, and plasma control issues

The eddy currents induced in three dimensional structures of in-vessel components in tokamaks and their influence on the plasma controls are investigated. The in-vessel components in ITER are used as an example. A simple method of modifying a toroidally symmetric two dimensional (2-D) model to reproduce the effects of three dimensional (3-D) in-vessel components is proposed. Several issues in plasma control are examined to determine whether the 2-D model adjusted by this method reproduces the results of the 3-D model. Analyses are performed for the plasma stabilization effects of in-vessel components, the evolutions of plasma position and shape in plasma disturbances, the magnetic probe signals and the stray fields at plasma breakdown. It is shown that this method of modifying 2-D models is acceptable for simulating plasma motions and probe signals. However, discrepancies between 3-D and 2-D models are observed in the probe signals during rapid motions of plasmas and stray fields at plasma breakdown. Plasma control issues, for which attention has to be paid to 3-D paths of eddy currents, are discussed

Beta enhancement of tokamak plasma with nearly circular cross-section

Access to the second stability region of the ballooning modes is obtained by optimizing the profiles of plasma pressure and safety factor in a conventional tokamak plasma. It is shown that access becomes easier at low shear and nearly circular cross-section. This result indicates the possibility of improving the beta limit of a tokamak without increasing the total plasma current or the elongation.

A matrix method for resistive MHD stability analysis of axisymmetric toroidal plasma

Abstract A matrix method to solve a resistive MHD stability problem has been developed. The equations are reduced to an eigenvalue problem of block tridiagonal matrices. The inverse iteration method is employed as a solution method of the eigenvalue problem.

Safety activities in JAERI related to ITER

Safety design approach, design codes and safety-related technology development for construction of the international thermonuclear experimental reactor (ITER) initiated in Japan Atomic Energy Research Institute (JAERI) has been reviewed. Safety design approach and implementation are focused on inherent and high-level passive safety features of the ITER-FEAT. By such features, escalation of abnormal condition to accident can be prevented without any special countermeasures such as emergency plasma shutdown system. As a result, the primary containment system and the secondary confinement one are proposed to achieve the ITER safety objectives. To establish the structural design codes for unique ITER components, an ad-hoc technical committee has been organized. The designs of the vacuum vessel and tritium process components have been reviewed and discussed. Recommendations including some attractive methodologies for structural design and components examinations, taking into account a low level of hazard potential and unique safety features of ITER was given by the committee. Another ad-hoc technical committee under JAERI Construction Division is discussing the design codes for ITER seismic isolation. The current status of R&D activities ongoing in JAERI for accumulation of experimental data required for assurance of the safety analysis and design codes for ITER are also reviewed.

Dynamic analysis of the tokamak support system in ITER

Abstract The ITER tokamak machine has a relatively flexible support structure against horizontal loads as from an earthquake. Modal analyses and modal transient analyses have been carried out by using the finite element analysis code NASTRAN to assess the response of the tokamak driven by a horizontal seismic load with the ITER design response spectra. Two types of three-dimensional models and a simplified mass/spring model have been used as structural models. The three-dimensional models have been used for modal analyses and modal transient analyses. The simplified mass/spring model has been used for parameter surveying of the transient response. It has been shown in the analyses that the lowest natural frequency of the tokamak is about 2 cycles s −1 and that the maximum relative response displacement between the magnet and the vacuum vessel is about 14 mm.

Convergence of solutions of the MHD stability code ERATO

Abstract Convergence of solutions of the ERATO code is studied numerically. Some irregularities of the computed data points on a convergence curve are observed in our investigation. It is conjectured that these irregularities are due to inaccuracy of the equilibrium or to inappropriate choice of the equilibrium calculation meshes.