Toroidal Field Ripple Reduction Research Articles

Active toroidal field ripple reduction system in FAST

The fusion advanced studies torus (FAST) has been proposed as a flexible and cost effective machine that is able to support the development of ITER and DEMO operating scenarios exploiting some innovative technology solutions and to investigate the physics of high-performance plasmas in a dimensionless parameter range close to ITER. The FAST magnet consists of 18 coils, spaced by 20° in the toroidal angle, each made up of 14 copper plates, suitably arranged in order to realise 3 turns in the radial direction, with 89.2 kA per coil (in the H-mode plasma scenario 6.5 MA at 7.5 T). The finite number and toroidal extension of the toroidal field coils (TFCs) cause a periodic variation of the toroidal field from its nominal value called toroidal field ripple (TFR). An active ripple reduction system has been comprehensively investigated, by using proper 3D finite elements models, to provide an efficient and flexible system able to minimize the TFR in the region of interest. An optimization study of position and size of the coils required to reduce to an acceptable level for the operations the maximum ripple on the plasma (well below 0.3%), feeding them with currents sustainable during the whole scenario (∼1/10 of the current flowing in TFCs), is presented in this paper.

Toroidal Field Ripple reduction studies for ITER and FAST

Two different approaches to control the Toroidal Field Ripple (TFR) amplitude in ITER and FAST devices are presented in this paper. The approach currently adopted to reduce the TFR in ITER is based on the installation of ferromagnetic inserts between the vacuum vessel shells. The same approach has been analyzed in the design of the Fusion Advanced Studies Torus (FAST) proposal. Details of the system's layout are given. A new approach based on the insertion of active coils between the outer legs of the Toroidal Field Coils (TFCs) and the plasma, has been extensively investigated for these two machines. This active system would allow reducing the TFR to values even smaller than with the ferromagnetic inserts. The case of a localized disturb like that introduced by a Test Blanket Module (TBM) for ITER is presented where only well localized active coils can produce a significant ripple reduction.

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

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 Alfvén velocity is found for suppressing RWM. High βN · HH98(y,2) of 2.2 with βN ∼ 2.3 and HH98(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) of ∼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 (∼0.8).

H-mode pedestal structure in the variation of toroidal rotation and toroidal field ripple in JT-60U

The effects of toroidal rotation and toroidal field (TF) ripple on the edge pedestal structure and edge-localized-modes (ELMs) were examined in JT-60U. The amplitude of the TF ripple was changed by the installation of ferritic steel tiles (FSTs). The profile of toroidal rotation VT became less counter with FSTs, particularly in the case where co-NBI was used in a large volume configuration. In this case, the plasma pressure was raised across the whole profile of the plasma. At the plasma edge, higher pedestal temperature was obtained with the growth of pedestal width. However, the effect of FSTs became less significant in a small volume configuration. As the VT became less counter at the pedestal, ELM frequency fELM was reduced and ELM energy loss ΔWELM was increased at fixed power crossing the separatrix Psep. The observed larger ELM energy drop at VT in a relatively co-direction involves the ELM perturbations of the electron temperature profile across an ELM that extends radially more inward. In addition, the inter-ELM transport loss is reduced and the pedestal pressure pped is weakly raised. The effect of FSTs appeared clearly in the large volume configuration where pped is raised even at a given VT at the pedestal while pped is not changed by FSTs at the small volume configuration. This suggests that the reduction of the TF ripple, the magnitude of which can affect the VT profile, plays a role in the increasing of pped. For increased pped due to the use of co-NBI and the reduction of TF ripple, the spatial width of the pedestal ion temperature became greater.

Preliminary Design and Analysis of ITER In-Wall Shielding

ITER in-wall shielding (IIS) is situated between the doubled shells of the ITER Vacuum Vessel (IVV). Its main functions are applied in shielding neutron, gamma-ray and toroidal field ripple reduction. The structure of IIS has been modelled according to the IVV design criteria which has been updated by the ITER team (IT). Static analysis and thermal expansion analysis were performed for the structure. Thermal-hydraulic analysis verified the heat removal capability and resulting temperature, pressure, and velocity changes in the coolant flow. Consequently, our design work is possibly suitable as a reference for IT's updated or final design in its next step.

Orbit Following Calculation of Energetic Ions in the Design of Ferritic Insertion in the JT-60U

The design process regarding ferritic insertion in the JT-60U is described from the viewpoint of the behavior of energetic ions. The confinement of energetic ions and the absence of unfavorable heat flux on the first wall were assessed using the Fully Three-Dimensional magnetic field OFMC code, which was developed for the toroidal field ripple reduction experiment program utilizing ferritic inserts in the JFT-2M. In the final design, the absorbed power in the neutral beam injection is improved by a factor of about 1.3 in a large volume plasma with Bt0 = 1.9 T.

Research Activities on Tokamaks in Japan: JT-60U, JFT-2M, and TRIAM-1M

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state operation. Plasma performance in transient and quasi steady states has been significantly improved, utilizing reversed shear and weak shear (high-βp) ELMy H-modes characterized by both internal and edge transport barriers and high bootstrap current fractions. Development of each key issue for advanced steady-state operation has also been advanced. Advanced and basic research of JFT-2M has been performed to develop high-performance tokamak plasma as well as the structural material for a fusion reactor. Toroidal field ripple reduction with ferritic steel plates outside the vacuum vessel is successfully demonstrated. No adverse effects to the plasma were observed with poloidal fields inside the vacuum vessel (partial covering). Preparation is in progress for full-scale testing of the compatibility of the ferritic steel wall (full covering) with plasma. A heavy ion beam probe has been installed to study H-mode plasmas. Compact toroid (CT) injection experiments are performed, showing deep CT penetration into the core region of the H-mode. The TRIAM project has investigated steady-state operation and high-performance plasma of a tokamak with the high toroidal magnetic field superconducting tokamak. Four important contributions in the fields of fusion technology of superconducting tokamaks, steady-state operation, high-performance plasma, and startup of plasma current without the assistance of center solenoid coils have been achieved on TRIAM-1M, especially regarding steady-state operation by realization of a discharge for >3 h.

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.

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.

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.

Advanced Material Tokamak Experiment (AMTEX) Program with Ferritic Steel for JFT-2M.

To examine the effects of ferritic steel on error field, ripple reduction and plasma properties (especially for improved confinements), Advanced Material Tokamak EXperiment (AMTEX) is planned for JFT-2M. The AMTEX program consists of the test of toroidal field ripple reduction with ferritic inserts and that of the compatibility between ferritic steel and the tokamak plasma. To reduce fast ion losses due to toroidal field ripple, the ripple amplitude and its toroidal mode number have to be reduced. The guideline of the design for the reduction in ripple amplitude (δ) by inserting Ferritic Boards (FB) is as follows; the wider and thicker FB should be positioned far from the plasma and be located just under the toroidal field coil. Elimination of a central part of the FBs is effective in reducing a second harmonic ripple amplitude. Since the ferritic steel easily saturates by the toroidal magnetic field, the δ depends on the toroidal magnetic field. The JFT-2M Vacuum Vessel (VV) will be remodelled using ferritic steel for the plasma-matching test. Since the relative magnetic permeability of the ferritic steel is around two, the effect of the ferritic steel VV on plasma equilibrium is rather small. A simple port design can reduce any ripple amplitude of the fundamental mode but cannot reduce that of higher modes.