In JFT-2M, ferritic steel plates (FPs) were installed on almost the whole inner surface of the vacuum vessel. This arrangement is called the ferritic inside wall (FIW), and is the third step of the advanced material tokamak experiment programme. The toroidal field (TF) ripple was reduced by optimizing the thickness of FPs but the total ripple structure has become more complex, with a non-periodic feature in the toroidal direction, because of the existence of ports and other components that limit the periodic installation of FPs. We investigated the effect of this complex ripple on the heat flux onto the first wall due to fast ion loss. The ripple trapped loss was reduced as a result of the reduced magnetic ripple of the FIW. Additional FPs were also installed outside the vacuum vessel to produce a localized larger ripple. A small ripple trapped loss was observed when the shallow ripple well exists in the poloidal cross section, and a large ripple trapped loss was observed when the ripple well extends deep into the plasma region. Experimental results were almost consistent with computation with a newly developed fully three-dimensional magnetic field orbit-following Monte-Carlo code including the three-dimensional complex structure of the TF ripple and the non-axisymmetric first wall geometry.

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Effects of complex magnetic ripple on fast ions in JFT-2M ferritic insert experiments

Recent JT-60U experimental results towards the development of a steady-state advanced tokamak scenario are presented. The JT-60U experiments within the past two years emphasize the extension of the operation regimes and the development of active control for high bootstrap current fraction plasmas with a strong linkage between pressure and current profiles. Reduction of a toroidal magnetic field ripple by installing ferritic steel tiles decreases the fast ion loss and consequently enables one to access the new regimes due to an increase in net heating power. The decrease in the fast ion loss also contributes to access to the new regimes through confinement improvement attributed to the reduction of counter toroidal rotation. High βN (∼4.2) exceeding the no-wall ideal limit is achieved at li = 0.8–1 in the high βp ELMy H-mode plasmas with a weak positive shear at the large volume configuration close to the wall. The large volume configuration had so far suffered from the large toroidal field ripple. The value of βN reaches the ideal wall limit, where RWM is suppressed by the plasma rotation. Small critical rotation velocity about 0.3% of Alfv´ en velocity is found for suppressing RWM. High βN · HH 98(y,2) of 2.2 with βN ∼ 2.3 and HH 98(y,2) ∼ 1 is sustained for 23.1 s, significantly longer than the current diffusion time (∼12τR) in the high βp ELMy H-mode plasmas. This duration is extended to a time scale for enhancement of recycling owing to a decrease in the wall-pumping rate. Successful recycling and density control is demonstrated for ∼30 s in the enhanced recycling regime even with outgas from the wall by maximizing the divertor pumping rate in the high-density ELMy H-mode plasmas. With small effects of the toroidal field ripple reduction, the controllability is investigated in the reversed shear (RS) plasmas with a high bootstrap current fraction (fBS )o f∼0.7–1. It is difficult to sustain these plasmas using a feedback control scheme for global parameter such as stored energy and the duration is determined by the ideal limit or slow ITB degradation in the linkage of current and pressure profiles. Real time qmin control is demonstrated with MSE diagnostic and LHCD at fBS = 0.46 with the strong pressure and current profile linkage. Physics studies also progress towards burning plasma control. Semi-global change of the ITB structure is observed in both high βp mode and RS plasmas. Complete NTM suppression is demonstrated with misalignment of ECCD position less than about half of the island width. Fast ELM propagation to the main chamber wall is observed, and significant hydrogen retention in the main chamber wall is found. In the plasma-shadowed area underneath the divertor region, the averaged carbon deposition rate is significantly low, however the (H + D)/C retention rate is high

https://iopscience.iop.org/article/10.1088/0029-5515/47/10/S06/pdf

Overview of JT-60U results for the development of a steady-state advanced tokamak scenario

The Japanese large tokamak JT-60 has been focusing its research emphasis on the development of high performance plasmas, with high confinement, high temperature and high density and non-inductive sustainment for a long time with possible minimization of external power input. The first demonstration of high bootstrap current discharges up to 80% of the total plasma current in a high-poloidal-beta (high-βp) mode and the concept development of a steady-state tokamak reactor (SSTR) based on this experimental achievement initiated the so-called 'advanced tokamak research'. The internal transport barriers (ITBs), discovered in the JT-60 high-βp mode, have been followed by worldwide explorations of reversed shear discharges associated with ITBs. The highest DT equivalent energy gain of was achieved in the JT-60 reversed shear H-mode discharges. The highest ion temperature of Ti = 45 keV and the highest fusion triple product of ni(0)τETi(0) = 1.5 × 1021 m−3 s keV were obtained in high-βp discharges. Advanced tokamak research is now the major trend of the current tokamak development. A new concept of compact ITER has been developed and proposed in the context of this advanced tokamak approach pursued on JT-60. Prospects for burning plasma physics have been investigated along the progress made in these modern tokamak experiments on JT-60 and related computer simulation analyses.

https://iopscience.iop.org/article/10.1088/0029-5515/45/8/026/pdf

Advanced tokamak research on JT-60

The iron-chromium alloy system has an unexplained anomaly: although there is a broad miscibility gap it appears to be favorable for chromium to dissolve in iron. This is consistent with ab initio calculation, but no simpler physically intuitive picture has been presented. Here it is shown that the Ising model, based on the bcc lattice with antiferromagnetic and ferromagnetic species, has the potential to exhibit similar behavior, with a skew miscibility gap arising from the solubility of antiferromagnetic species on nonadjacent sites. Essential characteristics of stainless steel (high Cr solubility and surface segregation) are correctly reproduced. Under some conditions, magnetization increases with temperature. The equilibrium miscibility gap due to mixed magnetism and segregation-driven positive dM/dT are fundamental features of the bcc Ising model itself, not just FeCr.

Magnetically induced immiscibility in the Ising model of FeCr stainless steel.

Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L–H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δv/v ~ 60% via non-resonant braking. Changes to global Δn/n, Δβ/β and ΔH98/H98 were ~3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

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ITER test blanket module error field simulation experiments at DIII-D

Compatibility between the plasma and low activation ferritic steel, which is a candidate material for fusion demonstration reactors, has been investigated step by step in the JFT-2M tokamak. We have entered the third stage of the Advanced Material Tokamak EXperiment (AMTEX), where the inside of the vacuum vessel wall is completely covered with ferritic steel plates ferritic inside wall (FIW). The effects of a FIW on the plasma production, impurity release, the operation region, and H-mode characteristics have been investigated. No negative effect has been observed up to now. A high normalized beta plasma of βN ~ 3, having both an internal transport barrier and a steady H-mode edge was obtained. A remarkable reduction in ripple trapped loss from 0.26 MW m−2 (without ferritic steel) to less than 0.01 MW m−2 was demonstrated by the optimization of the thickness profile of FIW. A code to calculate fast ion losses, taking into account the full three-dimensional magnetic structure was developed, and values obtained using the code showed good agreement with experimental results. Thus, encouraging results are obtained for the use of this material in the demo-reactor.

/pdf/high-performance-tokamak-experiments-with-a-ferritic-steel-vqkeivvsit.pdf

High performance tokamak experiments with a ferritic steel wall on JFT-2M

A new diagnostic method is proposed for direct measurements of electromagnetic field patterns of fast waves in toroidal plasmas. The fluctuations of the ponderomotive potential at the beat wave frequency between two fast waves with different frequencies produces the actual potential fluctuations via electron and ion motions. A heavy ion beam probe detected the potential fluctuation patterns at the beat wave frequency during the fast wave pulse. The amplitude of the potential fluctuations decreases with increasing electron temperature in fast wave current drive experiments in JFT-2M, owing to the improvement of wave absorption via electron Landau damping. The measured potential fluctuations do not contradict the prediction using full wave code calculations.

K. Kikuchi

Papers

High performance tokamak experiments with a ferritic steel wall on JFT-2M

A new diagnostic method for electromagnetic field patterns of fast waves during FWCD experiments in JFT-2M