Minority Heating Research Articles

The Ion Cyclotron, Lower Hybrid and Alfven Wave Heating Methods

This lecture covers the practical features and experimental results of the three heating methods. The emphasis is on ion cyclotron heating. First, we briefly come back to the main non-collisional heating mechanisms and to the particular features of the quasilinear coefficient in the ion cyclotron range of frequencies (ICRF). The specific case of the ion-ion hybrid resonance is treated, as well as the polarisation issue and minority heating scheme. The various ICRF scenarios are reviewed. The experimental applications of ion cyclotron resonance heating (ICRH) systems are outlined. Then, the lower hybrid and Alfvén wave heating and current drive experimental results are covered more briefly. Where applicable, the prospects for ITER are commented.

Production and diagnosis of energetic particles in FAST

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite facility to study fast-ion physics in deuterium plasmas under conditions relevant to a burning plasma. Energetic minority ions (H or 3He) accelerated by ion cyclotron resonance heating (ICRH), with dimensionless parameters close to those of fusion-born alphas in ITER, can be produced in FAST via 30 MW power ICRH minority heating. This work provides a first assessment of the extent to which the 3He fast-ion population can be diagnosed in FAST with a set of dedicated diagnostics for confined fast particles. Neutron emission spectroscopy (NES), gamma-ray spectroscopy (GRS) and collective Thomson scattering (CTS) diagnostics have been reviewed with a description of the state-of-the-art hardware and a preliminary analysis of the required lines of sight. The results of the analysis, based on numerical simulations of the spatial and energetic particle distribution function of the ICRH-accelerated ions for the standard FAST H-mode scenario, suggest that NES and GRS measurements can provide information on the fast 3He population effective tail temperature, with time resolutions in the range 20–100 ms. The proposed CTS diagnostic can measure the fast-ion parallel and perpendicular temperature with a spatial resolution of 5–10 cm and a time resolution of 10 ms. The paper provides a scientific basis for the prediction of the production and diagnosis of energetic ions in FAST.

Comparison of various wall conditionings on the reduction of H content and particle recycling in EAST

Reductions in H content and particle recycling are important for the improvement of ion cyclotron range of frequency (ICRF) minority heating efficiency and the enhancement of plasma performance of the EAST superconducting tokamak. During recent years several techniques of surface conditioning such as baking, glow discharge cleaning/ICRF discharge cleaning, surface coatings, such as boronization, siliconization and lithium coating, have all been attempted in order to reduce the H/(H+D) ratio and particle recycling in EAST. Even though boronization and siliconization were both reasonably effective methods to improve plasma performance, lithium coatings were observed to reduce the H content and particle recycling to levels low enough to allow the attainment of enhanced plasma parameters and operating modes on EAST. For example, by accomplishing lithium coating using either vacuum evaporation or the real-time injection of fine lithium powder, the H/(H+D) ratio could be routinely decreased to about 5%, which significantly improved ICRF minority heating efficiency during the autumn campaign of 2010. Due to the reduced H/(H+D) ratio and lower particle recycling, and a reduced H-mode power threshold, improved plasma confinement and the first EAST H-mode plasma were obtained. Furthermore, with increasing accumulation of deposited lithium, several new milestones of EAST performance, such as a 6.4 s-long H-mode, a 100 s-long plasma duration and a 1 MA plasma current, were achieved in the 2010 autumn campaign.

RF plasma production and heating below ion-cyclotron frequencies in Uragan torsatrons

In the IPP-Kharkiv there are two torsatrons (stellarators) in operation, and in both of them Alfvén resonance heating under high-k∥ conditions is used. This method of heating is advantageous for small-size devices, since in contrast to the minority and second-harmonic heating it can be realized at lower plasma densities. A series of experiments has been performed at the Uragan-3M torsatron with an aim to investigate the features of the discharge with a three-half-turn antenna. Electron temperatures in the range are achieved at plasma densities . The plasma energy content has increased by a factor of 2 with respect to the plasma produced with the frame antenna. A new four-strap shielded antenna has been manufactured and installed in the Uragan-2M. A high-frequency discharge for wall conditioning is introduced in the Uragan-2M torsatron. The discharge is sustained by a specially designed small frame antenna, and efficient hydrogen dissociation is achieved. A self-consistent model has been developed for simulation of plasma production in ICRF. The model includes a set of particle and energy-balance equations for the electrons, and the boundary problem for the Maxwell equations. The first calculation results on RF plasma production in the Uragan-2M stellarator with the frame-type antenna are presented.

Heat flux calculation and problem of flaking of boron carbide coatings on the Faraday screen of the ICRH antennas during Tore Supra high power, long pulse operation

Reliable and repetitive high power and long pulse tokamak operation is strongly dependant of the ability to secure the Plasma Facing Components (PFCs). In Tore Supra, a network of 7 infrared (IR) video cameras is routinely used to prevent PFCs overheating and damage in selected regions. Real time feedback control and offline analysis are essential for basic protection and understanding of abnormal thermal events. One important limitation detected by the IR real time feed-back loop during high power RF operation (injected power of 9.5 MW over 26 s and 12 MW over 10 s have been achieved respectively in 2006 and 2008) is due to the interaction between fast ions which increase the power flux density and flaking of the boron carbide coatings on the Faraday screen box of the ICRH antennas. An IR-based experimental procedure is proposed in order to detect new flakes during plasma operation. The thermal response of the B 4C coating is studied with and without flaking during plasma operation. The experimental heat flux deposited by fast ion losses on the Faraday screen is calculated for high (3.8 T) and low magnetic field (2 T) during high RF power operation (with fundamental hydrogen minority and second harmonic ICRH heating schemes respectively). The paper addresses both thermal science issues applied to machine protection and limitation due to fast ions issues during high RF power, long pulse operation. Safety margin to critical heat flux and number of fatigue cycles under heat load are presented in the paper.

ICRF mode conversion flow drive on Alcator C-Mod**2010 23rd IAEA Fusion Energy Conf. (Daejeon, Republic of Korea, 11–16 October) presentation EXW/4-1.

We have carried out a detailed study of ion cyclotron range of frequency (ICRF) mode conversion (MC) flow drive on the Alcator C-Mod tokamak including its dependence on plasma and RF parameters. The flow drive efficiency is found to strongly depend on the 3He concentration in D(3He) plasmas, a key parameter separating the ICRF minority heating regime and MC regime. At +90° antenna phasing (waves in co-Ip direction) and dipole phasing (waves symmetrical in both directions), we find that ΔV0, the change in the core toroidal rotation velocity, is in the co-Ip direction, increases with RF power and with Ip (opposite to the 1/Ip intrinsic rotation scaling). The flow drive efficiency decreases at higher plasma density and also at higher antenna frequency. The observed flow drive efficiency in H-mode has been small due to the unfavourable density scaling. The flow drive effect at −90° phasing appears to be saturated or decrease at high RF power. The up–down asymmetry in the MC to the ion cyclotron wave may be the key to understand the flow drive mechanism.

Ion cyclotron resonance heating with consistent finite orbit widths and anisotropic equilibria

Minority ion cyclotron resonance heating is studied using the self-consistent numerical model SCENIC. This model includes 3D geometries with full shaping and anisotropic pressure effects, warm contributions to the dielectric tensor and full orbit effects. It evolves the equilibrium, wave field and hot particle distribution function iteratively until a self-consistent solution is found. We will show applications to JET-like two-dimensional equilibria with minority heating scenarios. The effects due to different heating locations on the hot particle distribution function, the hot dielectric tensor and the equilibrium will be studied for symmetric wave injection. Finally, the RF-induced particle pinch is investigated using asymmetric wave injection.

Recent Progress in Plasma Control Studies on the Improvement of Plasma Performance in Heliotron J

Recent progress in plasma control studies on the improvement of plasma performance in Heliotron J is reviewed. The supersonic molecular beam injection (SMBI) fueling is successfully applied to Heliotron J plasma. A supersonic H2-beam is effectively injected to increase fueling efficiency and generate a peaked density profile. Local fueling with a short-pulsed SMBI can increase the core plasma density and avoid the degradation arising from edge cooling. Second harmonic electron cyclotron current drive (ECCD) experiments were conducted by launching a focused Gaussian beam with a parallel refractive index of −0.05 ≤ N‖ ≤ 0.6. Results show that the electron cyclotron (EC) driven current is determined not only by N‖ but also by local magnetic field (B) structure where the EC power is deposited. Detailed analysis of the observed N‖ and B dependences is in progress with a ray-tracing simulation using the TRAVIS code. Fast ion velocity distribution was investigated using fast protons generated by ion cyclotron resonant frequency (ICRF) minority heating. For the standard configuration in Heliotron J, charge exchange neutral particle analysis (CX-NPA) measurements show higher effective temperature of fast minority protons in the on-axis resonance case compared to that in the HFS (high field side) off-axis resonance case. However, the increase in bulk ion temperature in the HFS resonance case is larger than that in the on-axis resonance.

Numerical Analysis of ICRF Minority Heating in Helitoron J

The effect of the magnetic configuration on fast-ion confinement is one of the most important topics for helical devices. Fast-ion velocity distributions have been investigated using ion cyclotron range of frequencies (ICRF) minority heating in Heliotron J with special emphasis on the effect of the toroidal ripple (bumpiness) of the magnetic field strength. In measurements of the fast-ion tail generated by ICRF minority heating, a high bumpiness configuration is found to be preferable for tail formation. However, the measurement area based on the line of sight of the fast-ion detector was restricted in this experiment. Due to the complexity of the magnetic field in Heliotron J, three-dimensional analysis is required to interpret the experimental results. Monte-Carlo simulations were performed. The calculation results agree well with the experimental results for high-energy tail formation. The effective temperature of minority protons was estimated.

Study on Cavity Modes for D(H) Ion Cycloton Range Frequency Heating Scenarios in EAST

For low single-pass absorption of ion cyclotron range frequency (ICRF) wave in the EAST plasma cavity modes are expected to be excited between the low field side (LFS) antenna and the hybrid cut-off layer. The toroidal spectrum for D(H) minority heating scenarios in EAST is modeled by using FELICE (finite elements ion cyclotron emulator), a full wave code based on plane-stratified geometry. The excitation of cavity modes is studied. The methods for suppressing cavity modes are also discussed, to increase the efficiency of minority ion heating.

Diffusion coefficient and convective velocity profiles of Ni-impurity in JET plasmas with ICRH

The influence of the RF heating on the Ni-impurity transport is studied from comparative analysis of two discharges in JET: the reference discharge #69808 without RF power and discharge #68383 with the maximum ICRF power of 8.3 MW which is applied to electrons in Hydrogen Minority Heating scheme. With this goal we use multi-fluid Weiland model with trace impurity approximation and input profiles of density and temperature from the two shots. For the most instable determined eigen modes are calculated diffusivity, convective velocity and peaking factor for the two shots, respectively.

Measurement of the ratio of hydrogen to deuterium at the KSTAR 2009 experimental campaign

The control of the ratio of hydrogen to the deuterium is one of the very important issues for ion cyclotron range of frequency (ICRF) minority heating as well as the plasma wall interaction in the tokamak. The ratio of hydrogen to deuterium during the tokamak shot was deduced from the emission spectroscopy measurements during the KSTAR 2009 experimental campaign. Graphite tiles were used for the plasma facing components (PFCs) at KSTAR and its surface area exposed to the plasma was about 11 m(2). The data showed that it remained as high as around 50% during the campaign period because graphite tiles were exposed to the air for about two months and the hydrogen contents at the tiles are not fully pumped out due to the lack of baking on the PFC in the 2009 campaign. The validation of the spectroscopy method was checked by using the Zeeman effects and the ratio of hydrogen to the deuterium is compared with results from the residual gas analysis. During the tokamak shot, the ratio is low below 10% initially and saturated after around 1 s. When there is a hydrogen injection to the vessel via ion cyclotron wall conditioning and the boronization process where the carbone is used, the ratio of the hydrogen to the deuterium is increased by up to 100% and it recovers to around 50% after one day of operation. However it does not decrease below 50% at the end of the experimental campaign. It was found that the full baking on the PFC (with a high temperature and sufficient vacuum pumping) is required for the ratio control which guarantees the efficient ICRF heating at the KSTAR 2010 experimental campaign.

Residual gas analysis for long-pulse, advanced tokamak operation

A shielded residual gas analyzer (RGA) system on Tore Supra can function during plasma operation and is set up to monitor the composition of the neutral gas in one of the pumping ducts of the toroidal pumped limited. This "diagnostic RGA" has been used in long-pulse (up to 6 min) discharges for continuous monitoring of up to 15 masses simultaneously. Comparison of the RGA-measured evolution of the H(2)/D(2) isotopic ratio in the exhaust gas to that measured by an energetic neutral particle analyzer in the plasma core provides a way to monitor the evolution of particle balance. RGA monitoring of corrective H(2) injection to maintain proper minority heating is providing a database for improved ion cyclotron resonance heating, potentially with RGA-base feedback control. In very long pulses (>4 min) absence of significant changes in the RGA-monitored, hydrocarbon particle pressures is an indication of proper operation of the actively cooled, carbon-based plasma facing components. Also H(2) could increase due to thermodesorption of overheated plasma facing components.

Progress in Steady-State Plasma Operation Using ICRF Heating on LHD

As one of the main objectives of operation of the Large Helical Device (LHD), long-pulse plasma discharge experiments have been carried out using ion cyclotron range of frequency (ICRF) heating power (mainly using a minority heating method). Discharges with electron density ne < 1019 m-3 and Te0 > 1 keV have been sustained with an ICRF heating power around ~1 MW and an electron cyclotron heating power of ~0.1 MW. The total injected heating energy exceeded 1.6 GJ. Many technological improvements were made before undertaking long-pulse plasma discharges, e.g., the installation of a steady-state high-rf power generator and a liquid stub tuner capable of real-time impedance matching. Over the past decade, the achieved pulse length has increased to 1 h. One of the keys to this success was dispersion of the local plasma heat load onto divertors, which was accomplished by cyclically sweeping the magnetic axis inward and outward. Eventually, the plasma terminated due to the penetration of impurities, which originated from the release of thin flakes on the divertor plates. The pulse length was extended by installing new divertor plates with better thermal conduction. A mode conversion heating scenario has been considered as an alternative to the minority ICRF heating scenario; the former may have advantages due to the lack of an ion cyclotron resonance layer in front of the antennas in the mode conversion case.

MeV-range fast ion losses induced by fishbones on JET

Energy and pitch angle resolved measurements of highly energetic (megaelectronvolt (MeV) range) suprathermal ions ejected from the plasma through interaction with fishbone oscillations are presented. The measurements are obtained with a 2D scintillator probe diagnostic installed on JET, which is designed to detect lost ions only above a certain energy threshold (Emin,D ∼ 200 keV). In the case reported here the lost ions are identified as fast protons which had been accelerated to high energies by ICRF minority heating. The energy of the lost protons (∼0.5–4 MeV) is approximately one order of magnitude higher than the energy of the injected beam ions (maximum 130 keV) driving the fishbone. Losses arriving at the probe are enhanced by about a factor 10–20 with respect to MHD-quiescent levels, and are found to increase quadratically with the fishbone amplitude. Using a number of simplifying assumptions, numerical simulations have been performed which combine the HAGIS, MISHKA and SELFO codes (where the distribution function predicted by SELFO has been validated against neutral particle analyser measurements). The losses are found to originate from orbit stochastic diffusion of trapped protons near the plasma boundary or from counter-passing protons deep in the plasma core, which transit under the influence of the fishbone into an unconfined trapped orbit. The simulations show further that the losses are of non-resonant type. The simulated energy and pitch angle distribution of the losses, the temporal behaviour of the losses during a fishbone cycle and the scaling of the losses with the fishbone amplitude are compared with experiment. The simulation results are mostly in broad agreement with experiment, but some of the predictions could not be reconciled with experiment using this model.

Neutron emission levels during the ITER zero-activation phase

In recent experiments at JET, a contribution to the neutron emission from reactions between beryllium and 3He, 4He and H has been identified. With the beryllium wall planned for ITER, this raises the question of possible neutron activation during the ITER zero-activation phase. Here, we estimate the neutron emission rates for various heating scenarios foreseen for this ITER phase using Monte Carlo simulations. The emission is seen to be strongly dependent on the scenario chosen and the assumptions involved. We find that fundamental minority heating can contribute on the scale of low temperature deuterium plasmas, depending on minority concentration and ion cyclotron resonance heating (ICRH) power applied. Harmonic ICRH leads to production of tails that can give rise to significant neutron emission rates, while rates from hydrogen beams will be near zero. Better knowledge of the zero-activation phase conditions, and more sophisticated ICRH codes, would be needed to give exact rate predictions. We conclude that rates from the so-called zero-activation plasmas will be significantly lower than expected for the DD or DT phases, but far from zero.

Optimization Study of ICRF Heating in the LHD and HSX Configurations

AbstractTwo global simulation codes, TASK/WM (a full wave solver) and GNET (a 5‐D drift kinetic equation solver), are combined to simulate the ICRF heating in the 3D magnetic configuration. The combined code is applied to the ICRF minority heating in the LHD configuration. An optimization of the ICRF heating is considered in changing the magnetic configurations and the resonance surfaces in the LHD plasmas using GNET code. It is found that the heating efficiency is improved about 30% with the heating power of 10MW in the optimized heating scenario from that of the present standard off‐axis heating scenario. Also the ICRF minority heating is studied in the HSX plasma and it is found that the ICRF heating of about 100kW is still effective to heat the plasma even ϱ /a ∼ 1/7.5 for tail ions (© 2010 WILEY‐VCH Verlag GmbH &amp; Co. KGaA, Weinheim)

Tritium minority heating with mode conversion of fast waves

A new ion-heating scenario in tokamak plasmas, based on cyclotron damping of ion Bernstein waves (IBWs) by tritium minority at the first ion cyclotron harmonic (i.e., ω=2ΩcT), is proposed. The IBWs are coupled by mode conversion of fast magnetosonic waves in a D-H(T) (tritium minority in hydrogen-deuterium) plasma. The mode conversion layer is located near the center of the plasma column as well as the resonant layer of the tritium minority. A possible scenario for the JET (Joint European Torus) tokamak [J. Wesson, JET Report No. 99, 1999], based on the present idea, has been analyzed by means of the numerical codes TORIC and SSFPQL (toroidal ion cyclotron and steady state Fokker–Planck quasilinear) [M. Brambilla, Nucl. Fusion 34, 1121 (1994); Plasma Phys. Controlled Fusion 41, 1 (1999)]. As a result, tritium ions are accelerated up to energies close to the peak value of the DT cross section and steady state breakeven condition (Q≈1.3) can be reached with 25% minority tritium concentration.

The Ion Cyclotron, Lower Hybrid and Alfven Wave Heating Methods

This lecture covers the practical features and experimental results of the three heating methods. The emphasis is on ion cyclotron heating. First, we briefly come back to the main non-collisional heating mechanisms and to the particular features of the quasilinear coefficient in the ion cyclotron range of frequencies (ICRF). The specific case of the ion-ion hybrid resonance is treated, as well as the polarisation issue and minority heating scheme. The various ICRF scenarios are reviewed. The experimental applications of ion cyclotron resonance heating (ICRH) systems are outlined. Then, the lower hybrid and Alfvén wave heating and current drive experimental results are covered more briefly. Where applicable, the prospects for ITER are commented.

Alfvén resonance heating in Uragan-2M torsatron

Uragan-2M is a medium-size torsatron with reduced helical ripples. This machine has the major plasma radius R = 1.7 m, the average minor plasma radius rp ≤ 0.24 m and the toroidal magnetic field B0 ≤ 2.4 T. The Alfven resonance heating in a high k‖ regime is advantageous for small size machines since it can be realized at smaller plasma densities than the minority and second harmonic heating. The Alfven resonance heating is examined numerically in the approximation of radially non-uniform plasma cylinder with identical ends. The numerical model for wave excitation and propagation accounts for the longitudinal electron thermal motion and the finite ion gyroradius which allow the model to treat correctly the propagation and damping of the kinetic Alfven wave in hot plasma. A compact antenna consisting of four loop elements is chosen to provide operation in a high k‖ regime. The major drawback of such an approach is the presence of plasma peripher y heating owing to unavoidable excitation of low k‖ Alfven resonances. Calculations show that, with the proper choice of heating regime, the periphery heating has an acceptable level and the major part of the power is deposited inside plasma column.