Inboard Limiter Research Articles

The KSTAR plasma facing components for 2010 operation

The KSTAR plasma facing components (PFCs) consist of inboard limiter, poloidal limiter, divertor, passive stabilizer and neutral beam armor. The main function of the PFCs is to define boundary of operating plasma and to protect the vacuum vessel and in-vessel components such as diagnostic components, in vessel control coil and several kinds of launchers for heating and current drive systems. The divertor is designed to enhance effective particle control to keep high quality plasma with various flexibilities in the shaping control for wide range of operational regime. The passive stabilizer that is made of CuCrZr alloy is designed to passively control the vertical position and MHD instabilities during operation as well as outer boundary of the plasma. Since fabrication has been started for all of the plasma facing components from middle of 2009, the inboard limiter, the divertor, and the passive stabilizer were successfully installed in the vacuum vessel, in turn. Moreover, one set of neutral beam armor and three strings of poloidal limiters were also installed according to the heating system that newly comes in 2010. All the PFCs tiles were baked to 200 °C and the PFC system showed no vacuum leakage and other mechanical troubles. In this paper, key features, fabrication, results of assembly, and baking of the KSTAR PFCs are summarized in detail.

Generation of initial closed flux surfaces by ECH at a conventional aspect ratio of R/a ∼ 3: experiments on the LATE device and JT-60U tokamak

Generation of initial closed flux surfaces by electron cyclotron heating (ECH) at a conventional aspect ratio of R/a ∼ 3 is investigated on the Low Aspect ratio Torus Experiment (LATE) device and further tested on the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) tokamak. In both experiments, a plasma current is initiated and increased by ECH under steady external fields composed of a toroidal field and a weak vertical field. In the LATE experiments, a movable inboard limiter is used to change the aspect ratio of plasmas up to R/a ∼ 3. The results show that the formation of closed flux surfaces is still possible up to R/a ∼ 3 while higher decay indices of vertical field are required as the aspect ratio increases. Similar current start-up discharges are performed on JT-60U, and a plasma current is initiated and increased up to 20 kA under a vertical field having a high decay index.

A shaped First Wall for ITER

The ITER First Wall is being redesigned to address a number of issues identified during the 2007 design review. One of the main improvements concerns the handling of parallel plasma heat loads. The design must be optimised for maximum leading edge protection with acceptable power flux distribution, which is achieved by shaping the First Wall panels. The conceptual design presented in the paper can accommodate both inboard and outboard limiter plasmas for a total power in the discharge of 7.5MW at 7.5MA and allows the abandonment of the original dedicated port limiters.

Comparison of scrape-off layer profiles in outboard-versus inboard-limited plasmas in Tore Supra

In the Tore Supra tokamak, the scrape-off layer profiles of discharges limited either on the inboard or on the outboard side are compared. Inboard-limited discharges are characterized by substantially longer SOL e-folding lengths of ion and electron temperatures and electron density measured near the top of the plasma, providing strong evidence that the ion and electron energy transport across the SOL is enhanced on the outboard side, similar to the particle transport. The parallel heat flux density extrapolated to the last closed flux surface is found to be inversely proportional to its e-folding length in the SOL. The outboard limiters thus receive higher and more peaked heat loads than the inboard limiters. These results are important for the optimization of the plasma start-up scenario and design of the first wall in ITER.

Evidence for a poloidally localized enhancement of radial transport in the scrape-off layer of the Tore Supra tokamak

Near-sonic parallel flows are systematically observed in the far scrape-off layer (SOL) of the limiter tokamak Tore Supra, as in many L-mode X-point divertor tokamak plasmas. The poloidal variation of the parallel flow has been measured by moving the contact point of a small circular plasma onto limiters at different poloidal angles. The resulting variations of flow are consistent with the existence of a poloidally localized enhancement of radial transport concentrated in a 30° sector near the outboard midplane. If the plasma contact point is placed on the inboard limiters, then the SOL expands to fill all the space between the plasma and the outboard limiters, with density decay lengths between 10 and 20cm. On the other hand, if the contact point lies on the outboard limiters, the localized plasma outflux is scraped off and the SOL is very thin with decay lengths around 2–3cm. The outboard radial transport would have to be about two orders of magnitude stronger than inboard to explain these results.

Overview of the Alcator C-Mod program

Research on the Alcator C-Mod tokamak has emphasized RF heating, self-generated flows, momentum transport, scrape-off layer (SOL) turbulence and transport and the physics of transport barrier transitions, stability and control. The machine operates with PRF up to 6 MW corresponding to power densities on the antenna of 10 MW m−2. Analysis of rotation profile evolution, produced in the absence of external drive, allows transport of angular momentum in the plasma core to be computed and compared between various operating regimes. Momentum is clearly seen diffusing and convecting from the plasma edge on time scales similar to the energy confinement time and much faster than neo-classical transport. SOL turbulence and transport have been studied with fast scanning electrostatic probes situated at several poloidal locations and with gas puff imaging. Strong poloidal asymmetries are found in profiles and fluctuations, confirming the essential ballooning character of the turbulence and transport. Plasma topology has a dominant effect on the magnitude and direction of both core rotation and SOL flows. The correlation of self-generated plasma flows and topology has led to a novel explanation for the dependence of the H-mode power threshold on the ∇B drift direction. Research into internal transport barriers has focused on control of the barrier strength and location. The foot of the barrier could be moved to larger minor radius by lowering q or BT. The barriers, which are produced in C-Mod by off-axis RF heating, can be weakened by the application of on-axis power. Gyro-kinetic simulations suggest that the control mechanism is due to the temperature dependence of trapped electron modes which are destabilized by the large density gradients. A set of non-axisymmetric coils was installed allowing intrinsic error fields to be measured and compensated. These also enabled the determination of the mode locking threshold and, by comparison with data from other machines, provided the first direct measurement of size scaling for the threshold. The installation of a new inboard limiter resulted in the reduction of halo currents following disruptions. This effect can be understood in terms of the change in plasma contact with the altered geometry during vertical displacement of the plasma column. Unstable Alfven eigenmodes (AE) were observed in low-density, high-power ICRF heated plasmas. The damping rate of stable AEs was investigated with a pair of active MHD antennae.

Two-dimensional diagnostic of edge plasma structure using a lithium beam probe in a compact helical system

A neutral lithium beam probe for two-dimensional diagnosis of edge plasmas has been designed and installed on the compact helical system. A lithium beam with an energy of 15keV and a current of 0.1mA is used. The spatial resolution is about 10mm, and the time response is about 10ms. The beam penetration depth is expressed in terms of the line integral density ⟨nel⟩, which is about 2×1018m−2. The beam injection angle can be varied and the observation point covers the edge and separatrix region of the helical diverter configuration. Two-dimensional electron density profiles for electron cyclotron heating and neutral beam injection (NBI) heated plasmas are obtained near and outside the last closed flux surface (LCFS). Analysis for two-dimensional density profile reconstruction indicates that significant amounts of surface plasma are confined outside the LCFS for NBI plasmas even though the ergodic layer is cut by the vacuum chamber wall (inboard limiter configuration). The usefullness of this new two-dimensional diagnostic in the edge region is demonstrated.

KSTAR edge probe diagnostics

Edge probe diagnostics (EPDs) for the KSTAR consists of fixed Langmuir probe arrays, movable and reciprocating Langmuir probe assemblies. The probe array is designed to measure plasma basic parameters and their poloidal profiles in the plasma-surface interaction regions such as an inboard limiter and divertors. The movable and the reciprocating probe assemblies are designed to measure spatial profiles of plasma parameters in the scrape-off layer region and near the X point, respectively. In this work, the current engineering design activities for the KSTAR EPDs, and technical issues from sample fabrications and their performance tests are presented.

Island divertor investigations on the W7-AS stellarator

The effectiveness of 10 newly installed control coils, which had already been implemented in preparation for the upcoming divertor experiments, has been demonstrated in several important areas. Most importantly we have been able to show that the control coils can be used to influence the ratio of parallel to cross-field transport within an island by nearly a factor 2. High recycling conditions have been attained at the inboard limiters for ι( a)=5/9 but there was no indication of detachment. Excellent data consistency has been achieved during the high density phase of the high recycling discharges between the saturation current I sat measured by the target Langmuir probes and the equivalent values derived from the new absolutely calibrated signal of the 2D H α CCD camera. Poloidal as well as toroidal asymmetries caused considerable problems when we tried to relate measurements of different edge/target diagnostic signals, taken at equivalent positions around the machine. These asymmetries could with high probability be attributed to toroidally non-symmetrically distributed components like the diamagnetic loop (module 5) or the MINISOX camera (module 3).

Core-edge studies with boundary island configurations on the W7-AS stellarator

Core, edge and scrape-off-layer plasma behaviour is studied principally under conditions of an a = 5/9 boundary island configuration - which is relevant for the upcoming W7-AS divertor campaign - but for now with ten inboard sector limiters. The major focus is on compatibility between good core confinement and attainment of high recycling at the limiter. At low input power Pin0.4 MW, operation at densities necessary to attain effective divertor action in the future invariably leads to a transition to the ELM-free H-mode accompanied by lower edge densities and increased core radiation until radiation collapse ensues. Thereby, enhancement factors in E of nearly two above the international stellarator confinement scaling are transiently achieved. The threshold density ethr, necessary to attain the H-mode increases with heating power, such that at 2 MW NBI heating power the H-mode is completely suppressed and peak densities at the limiter exceeding 1.5 × 1020 m-3 are realized. The efficacy of newly-installed control coils designed to manipulate the island geometry is tested. Their influence on the core plasma is verified. Due to geometrical effects associated with the mutual shadowing of the inboard limiters, statements regarding the influence on island physics must await the divertor configuration.

Carbon-based materials thermal development testing and selection process for first wall and divertor applications

Carbon-based materials are being investigated for use as plasma facing surfaces in the burning plasma experiment (BPX) fusion device. Specifically, carbon-carbon (C-C) composites have been proposed in designs for both inboard limiter and divertor surfaces. This paper describes a development program to test and evaluate candidate composites and to select the composite type most suitable for the BPX requirements. Tests were formulated to examine the effects of composite architectures (fiber type, fiber placement and volume fraction, and matrix fraction) and composite processing on composite conductivity. Specimens of 1D, 2D, and 3D C-C composite architectures were tested in the High Temperature Materials Laboratory at Oak Ridge National Laboratory using an automated laser flash system to determine the thermal diffusivity of these candidate materials. Costs were considered as part of the selection process. Results indicate that a relatively inexpensive, low conductivity, multidimensional fiber composite will meet design requirements for the inboard limiter design. The high heat flux in the divertor region necessitates further material development to achieve a material that will meet this region's performance requirements.

First Wall Inboard Limiter Design Development for the Compact Ignition Tokamak

The purpose of the inboard limiter of the Compact Ignition Tokamak (CIT) is to protect the vacuum vessel from high heat loads during operation. The limiter components must be maintained remotely; therefore, they must be designed to reduce maintenance time as much as possible. This drives the design toward larger tiles to reduce the number of items to be replaced; however, other factors drive the design toward smaller tiles. The thickness of the tiles must also be as small as possible to reduce overall build of the distance from the inner magnet leg to the plasma. This paper presents the factors involved in sizing of the tiles and shows design concepts for several sizes of tiles. It also discusses the material development that has been accomplished to date and the development in work to increase the thermal conductivity and strength of carbon-carbon materials for use as the limiter material.

Guard limiter design for the first wall

The stainless steel first wall of the Fusion Experimental Reactor (FER) needs protection against melting from the intensive heat load released during a plasma disruption. The guard limiter is proposed as an innovative first wall protection concept. The guard limiter is located slightly inside the plasma chamber beyond the first wall surface and it covers part of the first wall. It is modularized and each module is equipped with its own structural support and cooling system so that the replacement of a module is easy and reliable. The guard limiter of the FER consists of 12 inboard and 12 upper modules corresponding to the number of torus sectors. The upper modules can also be used as a ramp-up limiter. Both inboard and upper guard limiters can be replaced with straight motion through the interval between the TF coils. The coverage of the inboard limiter is only about 20% because its toroidal width is restricted by the narrow distance between TF coils, while a larger coverage is possible for the upper limiters. Replacement with vacuum-tight operation is possible for the inboard limiter by using a vacuum flask design.

Numerical Simulation of the Plasma Current Quench Following a Disruptive Energy Loss

The plasma/electromagnetic interaction with poloidal field coils and nearby passive conductor loops during the current quench following a disruptive loss of plasma energy is simulated. By solving a differential/algebraic system consisting of a set of circuit equations (including the plasma circuit) coupled to a plasma energy balance equation and an equilibrium condition, the electromagnetic consequences of an abrupt thermal quench are observed. Limiters on the small and large major radius sides of the plasma are assumed to define the plasma cross section. The presence of good conductors near the plasma and a small initial distance (i.e., 5 to 10% of the plasma minor radius) between the plasma edge and an inboard limiter are shown to lead to long current decay times. For a plasma with an initial major radius R0 = 4.3 m, aspect ratio A = 3.6, and current Ip = 4.0 MA, introducing nearby passive conductors lengthens the current decay from milliseconds to hundreds of milliseconds.