Ion cyclotron range of frequency heating experiments on the large helical device and high energy ion behavior (original) (raw)
Related papers
Ion cyclotron resonance frequency heating in JET during initial operations with the ITER-like walla)
Physics of Plasmas, 2014
In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall material on the JET Ion Cyclotron Resonance Frequency (ICRF) operation was assessed and also the properties of JET plasmas heated with ICRF were studied. No substantial change of the antenna coupling resistance was observed with the ILW as compared with the carbon wall. Heat-fluxes on the protecting limiters close the antennas quantified using Infra-Red (IR) thermography (maximum 4.5 MW/m 2 in current drive phasing) are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can well reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. Some experimental facts indicate that main-chamber W components could be an important impurity source: the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions; the W content is also increased in ICRF-heated limiter plasmas; and Be evaporation in the main chamber results in a strong and long lasting reduction of the impurity level. The ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 20%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating efficiency; The ICRF power can be deposited at plasma centre and the radiation is mainly from the outer part of the plasma. Application of ICRF heating in H-mode plasmas started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core could be observed.
Status of the ITER ion cyclotron heating and current drive system
AIP Conference Proceedings, 2015
The paper reports on latest developments for the ITER Ion Cyclotron Heating and Current Drive system: imminent acceptance tests of a prototype power supply at full power; successful factory acceptance of candidate RF amplifier tubes which will be tested on dedicated facilities; further design integration and experimental validation of transmission line components under 6MW hour-long pulses. The antenna Faraday shield thermal design has been validated above requirements by cyclic high heat flux tests. R&D on ceramic brazing is under way for the RF vacuum windows. The antenna port plug RF design is stable but major evolution of the mechanical design is in preparation to achieve compliance with the load specification, warrant manufacturability and incorporate late interface change requests. The antenna power coupling capability predictions have been strengthened by showing that, if the plasma scrape-off layer turns out to be steep and the edge density low, the reference burning plasma can realistically be displaced to improve the coupling.
Japanese Journal of Applied Physics, 2004
Particles under the influence of the ion cyclotron range of frequency (ICRF) electromagnetic field were analyzed in the large helical device (LHD) by numerically solving the equation of motion instead of the guiding-center equation. Behaviors of the ICRF-heated particles in three cases of slowing-down by thermal electrons were compared. We also compared the characteristics of the ICRF-heated particles in the standard magnetic configuration (R ax ¼ 3:75 m) with those in the inwardly shifted magnetic configuration (R ax ¼ 3:6 m). It was found that the maximum energies of particles starting from the core plasma region exceed 300 keV and that such particles are confined within the vacuum vessel wall for 10 À4 s. It was confirmed that the ICRF-heated particles with energies of around 400 keV are lost through the divertor field lines. The maximum energy of the ICRF-heated particles starting from the core region rises with increasing electron temperature. As a result, the energy level relevant to the proton (p)-boron (11 B) fusion reaction (' 650 keV) was obtained when T e > 30 keV. Through acceleration in both the parallel and the perpendicular direction to the magnetic field, the high-energy chaotic orbit particles were produced by ICRF heating. It was also found that the energy at which the transition to the high-energy chaotic orbit particle occurs determines the upper energy limit of the ICRF-heated particle in LHD. A high confinement performance was found for the high-energy chaotic orbit particles produced by the ICRF field. The particle orbits in the inwardly shifted magnetic configuration were more widespread within the core region. Thus, it was concluded that the ICRF heating efficiency of the core plasma region in the inwardly shifted magnetic configuration exceeds that in the standard magnetic configuration.
Status of the ITER Ion Cyclotron H&CD
EPJ Web of Conferences, 2017
The ITER Ion Cyclotron Heating and Current Drive system (IC H&CD) is designed to deliver 20MW to a broad range of plasma scenarios between 40 and 55MHz, during very long pulses. It consists of two broadband equatorial port plug antennas, their pre-matching and matching systems, transmission lines, Radio Frequency (RF) Sources and High Voltage Power Supplies. The overall project schedule has been revised and agreed by ITER Council; it re-integrates the second antenna and its power supplies in construction baseline and sets the dates for progressive installation with DT phase planned in 2035. Recent progress on ICRF subsystems is reported, covering design evolution, qualification of test articles and specific R&D results in domestic agencies, suppliers, associated laboratories and IO.
ADX: a high field, high power density, advanced divertor and RF tokamak
Nuclear Fusion, 2015
The MIT PSFC and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX) [1]-a tokamak specifically designed to address critical needs in the world fusion research program on the pathway to DT fusion devices: 1. Demonstrate robust divertor power handling solutions at reactor-level boundary plasma parameters (heat fluxes, plasma pressures and PMI flux densities), which scale to long-pulse operation 2. Demonstrate nearly complete suppression of divertor material erosion, sufficient to sustain divertor lifetime for ~5x10 7 s of plasma exposure at reactor-level parameters 3. Achieve the above two goals while demonstrating a level of core and pedestal plasma performance that projects favorably to a fusion power plant and in physics regimes that are prototypical 4. Demonstrate efficient radio frequency current drive and heating techniques that solve plasma-material interaction challenges, scale to long-pulse operation and project to effective current profile control 5. Determine high-temperature PMI response of reactor-relevant plasma-facing material candidates, such as tungsten and liquid metals, in an integrated tokamak environment, assessing issues of material erosion, damage, material migration and fuel retention at reactor-level performance parameters. ADX is a high field (≥ 6.5 tesla, 1.5 MA), high power density facility (P/S ~ 1.5 MW/m 2) specifically designed to test innovative divertor ideas at reactor-level plasma/atomic physics parameters-divertor target plate conditions (e.g., T t < ~5eV, n t > ~10 21 m-3 [2]), boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region-while simultaneously producing high performance core plasma conditions prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fueling from external heating and current drive systems. Equally important, the experimental platform is specifically designed to test innovative concepts for lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side-the latter being a location where energetic plasma-material interactions can be controlled and favorable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination-advanced divertors, advanced RF actuators, reactorprototypical core plasma conditions-will enable ADX to explore integrated solutions compatible with attaining enhanced core confinement physics, such as made possible by reversed central shear and flow drive, using only the types of external drive systems that are considered viable for a fusion power plant. Critical need-solution for heat exhaust: As stated in 2013 EFDA report [3]: "A reliable solution to the problem of heat exhaust is probably the main challenge towards the realisation of magnetic confinement fusion. The risk exists that the baseline strategy pursued in ITER cannot be extrapolated to a fusion power plant. Hence, in parallel to the programme in support of the baseline strategy, an aggressive programme on alternative solutions for the divertor is necessary. Some concepts are already being tested at proof-of-principle level and their technical feasibility in a fusion power plant is being assessed. Since the extrapolation from proof-of-principle devices to ITER/DEMO based on modelling alone is considered too large, a dedicated test on specifically upgraded existing facilities or on a dedicated Divertor Tokamak Test (DTT) facility will be necessary." Critical need-solution for efficient steady state current drive/heating: In addition, current drive and heating technologies (and ideally flow drive) must achieve high system efficiency (wall-plug to plasma), high system availability, and operate reliably and continuously in a thermonuclear PMI environment; otherwise the tokamak concept as the basis for a steady-state, electricity producing power plant is at the risk of being a dead end.
Development of high current pulsed H-ion source and ECR ion source for the injector Linac at RRCAT
Development of a high current Hion source to serve as an injector for the front end H -Linac operating at 50 keV energy and 30 mA current has been initiated at Raja Ramanna Centre for Advanced Technology (RRCAT), Indore. A prototype filament driven multi-cusp Hion source with three-electrode ion extraction system has been designed, fabricated and undergoing trial runs of operation at ≤10 keV energy to generate ion beam current ≤1 mA. In the first phase of the plan, development of front end linac operating at 3 MeV energy and 30 mA current along with prototype fabrication of DTL and SFDTL structures, is envisaged. We report here on the progress made in the recent past towards development of a prototype filament driven multi-cusp Hion source. After completion of physics design of multi-cusp magnetic field using 12 Nd-Fe-B permanent magnets the designed parameters have been verified with the experimental results obtained using a 3D Hall-probe system used for field mapping inside the cylindrical plasma chamber. Placement of tungsten filament is done in the central field free region in triple hair-pin and helical shape. The low temperature, low density plasma generated in the multi-cusp field through filament heating has been characterized using a Langmuir probe technique. The ion beam extraction has been performed using three-electrode extraction geometry. Initial test performed using 5 kV field between plasma electrode and ground electrode has resulted in generation of 0.75 mA of hydrogen ions beam current. Further studies will be carried out after deployment of filter magnets and electron steering magnets to remove co-extracted electrons and test the Hion beam. We have attempted to perform generation of stable hydrogen ions beam using microwave generated CW ECR plasma source under solenoid fields in resonance conditions. Hydrogen ions beam of 7.8 mA has been extracted at 25 kV accelerating field using threeelectrode flat shape ion extraction geometry. The extraction geometry is being optimized using IGUN code with Pierce geometry angle between the plasma and puller electrodes with suitable gap between the electrodes to deliver high perveance and high brightness ion beam. Efforts will be made to operate ECR ion source as pulsed H-ion source, suitable for injector linac.