Core Confinement Research Articles

Anomalous modal structure in a waveguide with a photonic crystal core

We analyze a dielectric waveguide with a photonic crystal core. Using constant frequency contour analysis, we show that the modal behavior of this structure is drastically different from that of a conventional slab waveguide. In particular, at a given frequency the lowest-order guided mode can have an odd symmetry or can have more than one nodal plane in its field distribution. Also, there exist several single-mode regions with a different modal profile in each region. Finally, a single-mode waveguide for the fundamental mode with a large core and strong confinement can be realized. All these behaviors are confirmed by our three-dimensional finite-difference time-domain simulations.

Tokamak operation with high-Z plasma facing components

Plasma operation with high-Z plasma facing components is investigated with regard to sputtering, core impurity contamination and scenario restrictions. A simple model based on dimensionless quantities for fuel and high-Z ion sources and transport to describe the high-Z concentration in the plasma core is introduced. The impurity release and further transport is factorized into the sputtering yield, the relative pedestal penetration probability and a core confinement enhancement factor. Since there are quite large uncertainties, in particular, in the sputtering source and the edge transport of high-Z impurities, very different scenarios covering a wide parameter range are taken into account in order to resolve the experimental trends. Sputtering of tungsten by charge exchange neutrals in the energy range 0.5–2 keV is comparable to the effect of impurity ion sputtering, while the impact of thermal fuel ions is negligible. Fast ions produced by neutral beam injection as well as sheath acceleration during ICR heating may cause considerable high-Z sources if the limiters on the low-field side have high-Z surfaces. The critical behaviour of the central high-Z concentration in some experimental scenarios could be attributed to edge and core transport parameters in the concentration model. The improved H-mode with off-central heating turns out to be the most critical one, since a hot edge is combined with peaked density profiles.

Transition phenomena and thermal transport properties in LHD plasmas with an electron internal transport barrier

Two types of improved core confinement were observed during centrally focused electron cyclotron heating (ECH) into plasmas sustained by counter (CNTR) and Co neutral beam injections (NBI) in the Large Helical Device. The CNTR NBI plasma displayed transition phenomena to the high-electron-temperature state and had a clear electron internal transport barrier, while the Co NBI plasma did not show a clear transition or an ECH power threshold but showed broad high temperature profiles with moderate temperature gradient. This indicated that the Co NBI plasma with additional ECH also had an improved core confinement. The electron heat transport characteristics of these plasmas were directly investigated using heat pulse propagation excited by modulated ECH. These effects appear to be related to the m/n = 2/1 rational surface or the island induced by NBI beam-driven current.

A plastic damage approach for confined concrete

There are many situations in which it is necessary to increase the capacity of structures in use. This need maybe either for a change of use or because the structures have suffered some damage or have shown little resistance in case of extreme loads such as earthquakes. The most common methods for repair and retrofit of reinforced concrete columns are concrete jacketing, steel jacketing and fiber wrapping. This last type of reinforcement has many advantages as it offers a high-strength, low-weight and corrosion-resistant jacket with easy and rapid installation. The reinforcement with composite materials improves shear and compression strength and ductility as a result of concrete core confinement. The present analytical and numerical ability to quantify the efficiency of fiber confinement is rather limited, especially with respect to ductility. A constitutive model that approximately reproduces the behavior of structural concrete elements under confinement is developed in this paper. The model allows the assessment of concrete columns and bridge piles repaired and/or reinforced with fiber reinforced composites (FRP). The model presented is a modification of an existing coupled plastic damage model. A new definition for the plastic hardening variable and a new yielding surface with curved meridians are proposed. Both improvements enable the adequate reproduction of concrete behavior in high confinement conditions. The comparison of numerical and experimental results shows the model capacity to simulate concrete behavior under triaxial compression conditions like the ones present in concrete columns confined with fiber reinforced composites.

Transport modelling and gyrokinetic analysis of advanced high performance discharges

Predictive transport modelling and gyrokinetic stability analyses of demonstration hybrid (HYBRID) and advanced tokamak (AT) discharges from the International Tokamak Physics Activity (ITPA) profile database are presented. Both regimes have exhibited enhanced core confinement (above the conventional ITER reference H-mode scenario) but differ in their current density profiles. Recent contributions to the ITPA database have facilitated an effort to study the underlying physics governing confinement in these advanced scenarios. In this paper, we assess the level of commonality of the turbulent transport physics and the relative roles of the transport suppression mechanisms (i.e. E × B shear and Shafranov shift (α) stabilization) using data for select HYBRID and AT discharges from the DIII-D, JET and AUG tokamaks. GLF23 transport modelling and gyrokinetic stability analysis indicate that E × B shear and Shafranov shift stabilization play essential roles in producing the improved core confinement in both HYBRID and AT discharges. Shafranov shift stabilization is found to be more important in AT discharges than in HYBRID discharges. We have also examined the competition between the stabilizing effects of E × B shear and Shafranov shift stabilization and the destabilizing effects of higher safety factors and parallel velocity shear. Linear and nonlinear gyrokinetic simulations of idealized low and high safety factor cases reveal some interesting consequences. A low safety factor (i.e. HYBRID relevant) is directly beneficial in reducing the transport, and E × B shear stabilization can dominate parallel velocity shear destabilization allowing the turbulence to be quenched. However, at low-q/high current, Shafranov shift stabilization plays less of a role. Higher safety factors (as found in AT discharges), on the other hand, have larger amounts of Shafranov shift stabilization, but parallel velocity shear destabilization can prevent E × B shear quenching of the turbulent transport, and only E × B suppression is achieved.

Helium exhaust experiments on JET with Type I ELMs in H-mode and with Type III ELMs in ITB discharges

An analysis of helium exhaust experiments on JET in the MkII-GB divertor configuration is presented. Helium is pumped by applying an argon frost layer on the divertor cryo pump. Measurement of the helium retention time, , is performed in two ways: by the introduction of helium in gas puffs and measurement of the subsequent decay time constant of the helium content, ; and by helium beam injection and measurement of the helium replacement time, . In ELMy H-mode, with plasma configuration optimized for pumping, is achieved, where is the thermal energy replacement time. For quasi-steady internal transport barrier (ITB) discharges, the achieved is significantly lower. The achieved helium recycling coefficient, confirmed by an independent measurement to be Reff ≈ 0.91, is the same in both scenarios. None of the discharges are dominated by core confinement. The difference in is instead due to the confinement properties of the edge plasma, which is characterized by Type I ELMs for the H-mode discharges studied, and Type III ELMs for the quasi-steady ITB discharges. This difference is quantified by an independent measurement of the ratio of the helium replacement time with a helium edge source to the energy confinement time.

Ion Cyclotron Resonance Heating on TEXTOR

The original work done on TEXTOR on ion cyclotron resonance heating of the plasma is reviewed. After a brief introduction outlining the principles, the radio-frequency (rf) system is described, with its substantial evolution during time. Then, the different physics results are reviewed. Ion cyclotron heating has been performed in a large number of scenarios and under a wide range of conditions. Aside from the various minority or mode-conversion scenarios, the interaction with beam ions and the possibility of controlling fast-ion diffusion with rf have been deeply investigated. Both the interaction with the wall or edge plasma and the impact on improved core confinement were studied in detail. Pioneering work was done to demonstrate efficient heating with unshielded antennas and plasma production in a tokamak by rf alone for plasma startup assistance or wall conditioning.

Pedestal Characteristics of H-Mode Plasmas in JT-60U and ASDEX Upgrade

The role of the pedestal structure in ELMy H-mode plasmas for the core energy confinement and for the ELM energy losses were investigated in two tokamak-type devices, JT-60U and ASDEX Upgrade. The confinement degradation seen at higher densities is attributed to the reduction of the pedestal temperature limited by the ELM activities and the stiffness of the temperature profiles. In high triangularity or impurity seeded H-modes, in which higher energy confinement is generally achieved, a higher pedestal temperature is obtained by improving the edge MHD stability or the density profile peaking, respectively. The upper bound of the ELM energy loss is characterized by the pedestal energy. The ELM energy loss can become smaller at fixed pedestal energy when the pedestal collisionality is raised. Raising the pedestal collisionality also enhances the inter-ELM perpendicular transport loss and reduces the ELM loss power fraction. In ASDEX Upgrade, the continuous pellet injection is shown to effectively mitigate ELM losses while preserving the core confinement quality. In JT-60U, the operation regime of grassy ELM is investigated and shown to achieve highly integrated performance of high energy confinement with small ELMs in the low collisionality regime.

Suppression of large edge localized modes in high confinement DIII-D plasmas with a stochastic magnetic boundary

Large 70Hz Type-I edge localized modes (ELMs) are converted into small 130Hz oscillations using edge resonant magnetic perturbations (RMPs) from a coil with currents ⩽0.4% Ip in double null DIII-D plasmas. When the RMP is properly phased with respect to the background field errors, all but a few isolated ELM-like events are suppressed. The impulsive pedestal energy loss ΔEELM/Δt1/2 to the scrape-of layer is reduced a factor of ⩾20 relative to the Type-I ELMs and the core confinement is unaffected by the perturbation field. Significant changes in the properties of the ELMs are also observed when edge RMPs are applied to lower single null plasmas but the nature of these changes are much more complex. Both lower single null and double null plasmas are being studied to determine how ELM control techniques based on the application of edge RMPs can be expected to scale to future devices such as ITER.

ITER-relevant H-mode physics at ASDEX Upgrade

A variety of techniques are being tested on ASDEX Upgrade for avoiding or reducing the divertor energy load caused by ELMs. Quiescent H-mode operation has been improved by operation with recently boronized vessel walls. Zeff values in QH-modes are now ∼2.5, comparable with those found in ELMy H-modes of similar density. Type II ELM operation has been generated in discharges with ICRF additional heating alone, thus demonstrating its compatibility with reactor-relevant low core momentum and particle input. Furthermore, Type II ELM operation at low (<1) values of edge collisionality has been achieved using strong plasma shaping and high input power. Modification of the ELM frequency and of the ELM energy has been demonstrated using pellet injection, supersonic pulsed gas injection, magnetic triggering via vertical position oscillations and modulated edge electron cyclotron heating and current drive. ELM control using pellet pace making is now being incorporated into integrated scenarios that aim to provide energy and particle exhaust compatible with ITER requirements while maximizing the core confinement.

Suppression of large edge-localized modes in high-confinement DIII-D plasmas with a stochastic magnetic boundary.

A stochastic magnetic boundary, produced by an applied edge resonant magnetic perturbation, is used to suppress most large edge-localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H mode transport barrier and core confinement are unaffected by the stochastic boundary, despite a threefold drop in the toroidal rotation. These results demonstrate that stochastic boundaries are compatible with H modes and may be attractive for ELM control in next-step fusion tokamaks.

High recycling steady H-mode regime in the JFT-2M tokamak

A new operating regime, ‘high recycling steady’ (HRS) H-mode regime, has been discovered in the JFT-2M tokamak. It is easily reproduced with neutral beam heating of any co-, balanced-, or counter-injection under wall fuelling from the boronized first wall. The HRS is characterized by good energy confinement (H89P ∼ 1.6) at a high density (ne/nGW ∼ 0.7; nGW is the Greenwald density), low radiation loss power fraction (, typically), and the complete disappearance of large ELMs. Accompanying the HRS H-mode transition, the coherent magnetic fluctuations in the frequency range of the order of 10–100 kHz with significant variation are seen on magnetic probes at the vessel wall. In addition, the HRS H-mode edge condition is compatible with an improved core confinement. We have demonstrated that an internal transport barrier can be produced under a HRS H-mode edge condition, achieving βNH89P ∼ 6 at ne/nGW ∼ 0.7–1.0, transiently.

Study of internal transport barriers in the initial phase of Ohmic discharges in TUMAN-3M

A regime with electron heat confinement improvement was recently found in the initial phase of discharges in the TUMAN-3M tokamak. An internal transport barrier (ITB) formation in this regime was confirmed by Thomson scattering measurements and by transport modelling. Two possible reasons for the ITB formation are discussed in the paper: by reduction of turbulent transport in the presence of low magnetic shear or by plasma sheared rotation. It is demonstrated that low magnetic shear formation is possible in the current ramp-up phase of the Ohmic discharge. The low magnetic shear does not seem to be the only reason for the transport reduction. Results of Doppler reflectometry measurements of poloidal rotation of density fluctuations are presented. It is found that core confinement improvement correlates with the appearance of sheared rotation of the density fluctuations and with a burst of the MHD activity. The ITB formation in the regime seems to be a result of a combined action of reduced magnetic shear and plasma sheared rotation.

Status of and prospects for advanced tokamak regimes from multi-machine comparisons using the ‘International Tokamak Physics Activity’ database

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses (‘hybrid’ scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, , is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

JET and the Physics Basis of ITER

JET has made unique contributions to the physics basis of ITER by virtue of its ITER-like geometry, large plasma size and D-T capability. The paper discusses recent JET results and their implications for ITER in the areas of standard ELMy H-mode, D-T operation and advanced tokamak modes. In ELMy H-mode the separation of plasma energy into core and pedestal contributions shows that core confinement scales like gyroBohm transport. High triangularity has a beneficial effect on confinement and leads to an integrated plasma performance exceeding the ITER Q = 10 reference case. A revised type I ELM scaling predicts acceptable ELM energy losses for ITER, while progress in physics understanding of NTMs shows how to control them in ITER. The D-T experiments of 1997 have validated ICRF scenarios for heating ITER/a reactor and identified ion minority schemes (e.g. (3He)DT) with strong ion heating. They also show that the slowing down of alpha particles is classical so that the self-heating by fusion alphas should cause no unexpected problems. With the Pellet Enhanced Performance mode of 1988, JET has produced the first advanced tokamak mode, with peaked pressure profiles sustained by reversed magnetic shear and strongly reduced transport. More recently, LHCD has provided easy tuning of reversed shear and reliable access to ITBs. Improved physics understanding shows that rational q-surfaces play a key role in the formation and development of ITBs. The demonstration of real time feedback control of plasma current and pressure profiles opens the path towards fully controlled steady-state tokamak plasmas.

Chapter 8: Edge and Divertor Physics in ASDEX Upgrade

An overview of edge and divertor physics research on ASDEX Upgrade of relevance for next-step fusion devices like ITER is presented. The results described were primarily obtained in lower single-null divertor configurations with three consecutive bottom divertor designs, starting from an initial open divertor (Div I) over the closed LYRA configuration (Div II), optimized for low-triangularity single-null equilibria, to the presently operational variant Div IIb, fitting a large variety of plasma shapes. The upper, geometrically open divertor structure remained essentially unchanged. A dedicated diagnostics system in combination with advanced plasma control scenarios and extensive numerical modeling allowed for a detailed analysis of edge and divertor physics mechanisms. Main chamber edge profiles exhibit a double structure, especially pronounced in high-performance H-mode plasmas. While radial transport inside and across the separatrix is governed by critical gradients, the cold scrape-off layer wing shows rapid diffusion or even outward drift, probably related to intermittent crossfield transport. The divertor behavior has been studied for the different divertor geometries and for all operational regimes of interest. Closed divertor operation enhances divertor recycling and pumping, reduces the power load on target plates by increased upstream losses, and facilitates onset of plasma detachment. The transient power load during type I ELMs, however, remains high and problematic, while the small type III ELMs, appearing, for example, in radiative discharge scenarios, and especially the type II ELMs are nearly invisible on the target heat flux. Despite this strong effect of divertor geometry on the divertor behavior, its direct effect on core confinement remains small.

Observation of High Recycling Steady H-mode edge and compatibility with improved core confinement mode on JFT-2M

A new operational regime has been discovered on JFT-2M under the boronized first wall condition to produce the High Recycling Steady (HRS) H-mode, which is characterized by good energy confinement (H89P ∼ 1.6) at a high density of around 70% of the Greenwald density (ne/nGW), low radiated power fraction, and complete disappearance of large ELMs. Accompanying the HRS H-mode transition, coherent magnetic fluctuation is seen on magnetic probes at the vessel wall. These coherent fluctuations have a frequency of the order of 10–100 kHz with significant variation. They are recognized as important in enhancing particle transport, whose characteristics are similar to the Enhanced Dα (EDA) H-mode regime reported from Alcator C-Mod. The H′-mode, previously observed on JFT-2M, had features in common with the EDA mode in terms of the coherent density fluctuation, but it appeared transiently. The HRS operating regime is also similar to the EDA mode, except that HRS is seen even at low q95 (2 < q95 ⩽ 3). The most important feature of the HRS H-mode edge condition is compatibility with an improved core confinement mode at a high density without large ELMs. We have demonstrated that an internal transport barrier can be produced under a HRS H-mode edge condition, achieving βNH89P ∼ 6.2 at ne/nGW ∼ 70%, transiently.

Effect of the q(r) profile on electron transport barrier formation in the T-10 tokamak

The effect of the q(r) profile on transport barrier formation has been investigated in the T-10 tokamak using rapid current ramp-up with electron cyclotron resonance heating (ECRH). The enhanced core confinement (formation of the internal transport barrier (ITB)) arises earlier than the additional current penetrates in the core region. It is suggested that this process takes place due to decrease of magnetic surface density in the region of the rational q = 2 surface due to βp decrease, which is equivalent to dq/dr decrease near q = 2 surface. At various q-profiles, internal and external barriers were observed. Transient confinement improvement inside the ITB after the ECRH switch-off was also observed.

Two-dimensional diagnostics of edge and divertor region of toroidal helical plasmas using a lithium beam probe

A two-dimensional neutral lithium beam probe has been designed and installed on the Compact Helical System (CHS) to study the edge and separatrix region of l=2 helical divertor configuration. A 10–20 keV beam with equivalent neutral beam current of 0.1 mA is injected. Light emission from the injected beam by electron impact excitation (LiI/670.8 nm) is collected by an eight-channel detector system using optical band-pass filters and Avalanche Photodiodes. The injection beam line is mechanically tilted to cover the area two-dimensionally. The total system has been calibrated by injecting the beam into the helium-gas-filled CHS vacuum chamber. Detailed measurements of two-dimensional structure of edge and divertor region are expected to contribute to studies of helical divertor concept and to improve understanding of the role of edge plasmas on core confinement as well.

ELM moderation with ICRF heating on JET

A high-density, ELM-free, `enhanced Dα' (EDA) H-mode has been demonstrated with Ohmic or RF heating on the Alcator C-Mod device. An allied `low particle confinement' H-regime was also observed previously in `Mk0' divertor experiments on JET. Initial attempts to recover similar states in the JET `MkIIGB' divertor configuration are reported, employing appropriately shaped magnetic equilibria. Comparisons between cases with and without divertor cryopumping show enhanced edge sources tend to increase ELM frequency and reduce their size, without degrading confinement. However, torus pressure remains below that in C-Mod owing to lower wall recycling in JET. With cryopumping on, ion cyclotron resonance heated (ICRH)-dominated combined heating may be adapted for maximum magnetohydrodynamic stability, leading to recurrent ELM-free H-mode phases. Core confinement peaks as density and edge radiation rise in each ELM-free phase, while only small compound ELMs occur between them. Moreover magnetic and density fluctuations in each quiescent period are concentrated around two frequencies, although these modes appear to rotate in the opposite direction to coherent modes seen during EDA on C-Mod. Pedestal collisionality also remains much lower in both latest and even earlier LPCH JET cases. Finally an effect of heating scheme itself on edge properties is indicated by alternative H-modes at the same density and power, with lower pedestal electron temperature and higher ELM frequency occurring for ICRH than for NBH. Closer matches in dimensionless plasma variables would clarify scalability of EDA-mode and its potential as an integrated next-step regime.