K. Tsuzuki

Initial physics achievements of large helical device experiments

The Large Helical Device (LHD) experiments [O. Motojima, et al., Proceedings, 16th Conference on Fusion Energy, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997), Vol. 3, p. 437] have started this year after a successful eight-year construction and test period of the fully superconducting facility. LHD investigates a variety of physics issues on large scale heliotron plasmas (R=3.9 m, a=0.6 m), which stimulates efforts to explore currentless and disruption-free steady plasmas under an optimized configuration. A magnetic field mapping has demonstrated the nested and healthy structure of magnetic surfaces, which indicates the successful completion of the physical design and the effectiveness of engineering quality control during the fabrication. Heating by 3 MW of neutral beam injection (NBI) has produced plasmas with a fusion triple product of 8×1018 keV m−3 s at a magnetic field of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.4 keV have been achieved. The maximum s...

Demonstration of ripple reduction by ferritic steel board insertion in JFT-2M

In the JFT-2M tokamak, the application of low activation ferritic steels to plasmas has been investigated (the so-called Advanced Material Tokamak Experiment (AMTEX) programme). In the first stage, toroidal field ripple reduction was examined by ferritic steel boards (FBs) inserted between the toroidal field coils and the vacuum vessel. It is demonstrated that FB insertion reduced toroidal field ripple and the losses of fast ions produced by tangential co-NBI. By optimizing the FB thickness, such that fundamental mode ripple is minimized to 0.07% at the shoulder of the inside wall, ripple trapped loss is reduced to an almost negligible level. It is determined that the reductions of the fundamental mode ripple and the ripple banana diffusion coefficients at the shoulder are most effective in reducing ripple ion losses. Ripple loss reduction by FBs is also confirmed with perpendicular beam injection. The insertion of FBs causes no undesirable effects on plasma production and control.

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

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.

Recent progress on the development and analysis of the ITPA global H-mode confinement database

This paper describes the updates to and analysis of the International Tokamak Physics Activity (ITPA) Global H-Mode Confinement Database version 3 (DB3) over the period 1994–2004. Global data, for the energy confinement time and its controlling parameters, have now been collected from 18 machines of different sizes and shapes: ASDEX, ASDEX Upgrade, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M, JT-60U, MAST, NSTX, PBX-M, PDX, START, T-10, TCV, TdeV, TFTR and TUMAN-3M. The database now contains 10382 data entries from 3762 plasma discharges, including data from deuterium–tritium experiments, low-aspect ratio plasmas, dimensionless parameter experiments and plasmas. DB3 also contains an increased amount of data from a range of diverted machines and further data at high triangularity, high density and high current. A wide range of physics studies has been performed on DB3 with particular progress made in the separation of core and edge behaviour, dimensionless parameter analyses and the comparison of the database with one-dimensional transport codes. The errors in the physics variables of the database have also been studied and this has led to the use of errors in variables fits. A key aim of the database has always been to provide a basis for estimating the energy confinement properties of next step machines such as ITER, and so the impact of the database and its analysis on such machines is also discussed.

Plasma confinement studies in LHD

The initial experiments on the Large Helical Device (LHD) have extended confinement studies on currentless plasmas to a large scale (R = 3.9 m, a = 0.6 m). Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 1018m-3keVs at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 1019 m-3. The maximum stored energy reached 0.22 MJ with neither unexpected confinement deterioration nor visible MHD instabilities, which corresponds to β = 0.7%. Energy confinement times reached a maximum of 0.17 s. A favourable dependence of energy confinement time on density remains in the present power density (~40 kW/m3) and electron density (3 × 1019 m-3) regimes, unlike the L mode in tokamaks. Although power degradation and significant density dependence are similar to the conditions on existing medium sized helical devices, the absolute value is enhanced by up to about 50% from the International Stellarator Scaling 95. Temperatures of both electrons and ions as high as 200 eV were observed at the outermost flux surface, which indicates a qualitative jump in performance compared with that of helical devices to date. Spontaneously generated toroidal currents indicate agreement with the physical picture of neoclassical bootstrap currents. Change of magnetic configuration due to the finite β effect was well described by 3-D MHD equilibrium analysis. A density pump-out phenomenon was observed in hydrogen discharges, which was mitigated in helium discharges with high recycling.

Experimental studies of the dynamics of compact toroid injected into the JFT-2M tokamak

We present the first results from recent compact toroid (CT) injection experiments in the JFT-2M tokamak using the improved CT injector and diagnostics with fast time resolution. We have observed that the core line density increases rapidly at a maximum rate of ~1.3 × 1022 m−3 s−1 after a delay of 100–200 µs. This increment rate of the core density is about several times larger than that obtained so far. Interferometry measurement along the peripheral line chord of R = 1.1 m in the inboard side indicates that CT plasma reaches a region near the plasma core beyond the separatrix. Time-frequency and space distribution analyses of edge magnetic probe signals show that the magnetic fluctuation induced by the CT has the spectral peak at 250–350 kHz and propagates in the toroidal direction at the Alfven speed of the order of 106 m s−1. These results indicate the excitation of Alfven wave by CT injection. We have observed that the fluctuation level of the ion saturation current in the divertor and the Dα spectral line intensity decrease significantly after CT injection. Corresponding increase in the soft x-ray signals in the core region may suggest that the CT causes a transition to H-mode-like discharges.

Progress of advanced material tokamak experiment (AMTEX) program on JFT-2M

Application testing of the low activation ferritic steel to plasma is in progress in the JFT-2M tokamak. The toroidal field ripple reduction with ferritic insertion between the vacuum vessel and the toroidal field coils (1st stage) was successfully demonstrated. So far, no deteriorating effects of ferritic boards (FBs) inside the vacuum vessel has been observed in the pre-testing on compatibility with plasma (2nd stage). The density limit was improved by more than a factor of 1.6 after boronization with inside FBs. Design and preparation works are in progress for the testing on compatibility with plasma (3rd stage), where inside wall of the vacuum vessel will be fully covered with ferritic steel.

Experiments on NBI plasmas in LHD

Neutral beam injection (NBI) heating started in the second experimental campaign of the Large Helical Device (September to December 1998) by two tangential beam lines. With 100 keV hydrogen, the beam port through power of up to 3.7 MW was injected for 1 s typically. The energy confinement was systematically better than that predicted by the International Stellerator Scaling 95 up to a factor of 1.5. The temperature pedestal observed contributes to this enhancement. We have also demonstrated a long pulse discharge by NBI in the LHD. By injecting 0.7 MW of beam, a plasma with a density of 0.3 × 1019 m-3 was sustained for 22 s. A unique oscillating phenomenon of plasma quantities with a long time scale was observed in the long pulse discharge. One of the topics of NB discharge is that the plasma can be started up by NB alone. This technique is unique for helical systems that have a vacuum magnetic field confining high energy ions, and it is useful for helical systems to be free from the constraint of magnetic field strength that must coincide with the frequency required by electron cyclotron resonance heating (ECH).

Design and first experimental results of toroidal field ripple reduction using ferritic insertion in JFT-2M

In order to test the effect of the toroidal field ripple on the fast ion losses, ferritic steel boards were inserted between the vacuum vessel and the toroidal field coil in the JFT-2M tokamak. The experimental and computational results show that the ripple amplitude is reduced from 2.2 to 1.1%. The ion losses are monitored from the increase in the wall temperature measured by the infrared TV. The region of the ripple trapped ion losses moves to outer side by about 6 cm and the increment of the wall temperature due to the ion loss of ripple trapped and banana drift is reduced. The ripple loss of the fast ions is reduced by ferritic steel insertion for the first time in the world. No deleterious effect of the ferritic insertion on plasma production and plasma control has been observed so far.

Scaling of the energy confinement time with β and collisionality approaching ITER conditions

The condition of the latest version of the ELMy H-mode database has been re-examined. It is shown that there is bias in the ordinary least squares regression for some of the variables. To address these shortcomings three different techniques are employed: (a) principal component regression, (b) an error in variables technique and (c) the selection of a better conditioned dataset with fewer variables. Scalings in terms of the dimensionless physics variables, as well as the standard set of engineering variables, are also derived. The new scalings give a very similar performance for existing scalings for ITER at the standard beta(n) of 1.6, but a much improved performance at higher beta n.