Magnetic Fusion Devices Research Articles

Influence of the Charge Exchange Reactions on Plasma Parameters in Magnetic Fusion Devices

In this study, we are going to construct a model for particle and power balance in zero-dimensional model. The new model is based on Hirooka four-reservoir (core, SOL, gas, and wall) model. To compare the model predictions with experimental data and test the validity of the model, we used the parameters of TRIAM and ITER tokamak. The qualitative agreement between the model and the experimental results has been presented. Furthermore, our simulation results illustrate that charge exchange reactions have been identified to play an important role in plasma parameters.

Temporal and spatial evolution measurement of laser-induced breakdown spectroscopy on hydrogen retention in tantalum

Fuel retention measurement on plasma-facing components is an active field of study in magnetic confinement nuclear fusion devices. The laser-induced breakdown spectroscopy (LIBS) diagnostic method has been well demonstrated to detect the elemental distribution in PFCs. In this work, an upgraded co-axis LIBS system based on a linear fiber bundle collection system has been developed to measure the hydrogen (H) retention on a tantalum (Ta) sample under a vacuum condition. The spatial resolution measurement of the different positions of the LIBS plasma can be achieved simultaneously with varying delay times. The temporal and spatial evolution results of LIBS plasma emission show that the H plasma observably expands from the delay times of 0–200 ns. The diameter of Ta plasma is about 6 mm which is much less than the size of H plasma after 200 ns. The difference in the temporal and spatial evolution behaviors between H plasma and Ta plasma is due to the great difference in the atomic mass of H and Ta. The depth profile result shows that H retention mainly exists on the surface of the sample. The temporal and spatial evolution behaviors of the electron excited temperature are consistent with that of the Ta emission. The result will further improve the understanding of the evolution of the dynamics of LIBS plasma and optimize the current collection system of in situ LIBS in fusion devices.

Low-power-threshold parametric decay instabilities of powerful microwave beams in toroidal fusion devices

We discuss the experimental conditions responsible for a drastic decrease in the power threshold of parametric decay instabilities under auxiliary electron cyclotron resonance heating (ECRH) in toroidal magnetic fusion devices when the upper hybrid (UH) resonance for the pump wave is absent. We show that for a finite-width pump in the presence of a nonmonotonic (hollow) density profile occurring due to plasma equilibrium in the magnetic islands or anomalous particle fluxes from the ECR layer, 3D localization of one or both daughter waves is possible. This localization leads to the full suppression of daughter wave energy losses from the decay layer and a substantial increase in the nonlinear pumping efficiency. This decreases the power threshold of nonlinear excitation, which can be easily overcome in current ECRH experiments utilizing 1 MW microwave beams. Different scenarios of extraordinary and ordinary wave decays are investigated. The secondary decays of primary daughter waves and pump wave depletion are considered as the most effective mechanisms leading to the transition of primary instability to the saturation regime. The proposed theoretical model was shown to be able to describe the anomalous phenomena discovered in ECRH experiments in different toroidal fusion devices all over the world.

Neutronics Assessment of a Compact D-D Neutron Generator as a Neutron Source for the Neutron Calibration in Magnetic Confinement Fusion Devices

Neutronics Assessment of a Compact D-D Neutron Generator as a Neutron Source for the Neutron Calibration in Magnetic Confinement Fusion Devices

Verification of an improved equation-free projective integration method for neoclassical plasma-profile evolution in tokamak geometry

A brute-force, long-time gyrokinetic simulation of plasma profile evolution in magnetic fusion devices is not desirable due to large computational resource requirements and a possible accumulation of numerical error. The equation-free projective integration method of Keverekidis et al. [Commun. Math. Sci. 1(4), 715–762 (2003)] is one of the outstanding candidates in projecting micro-scale simulations to a longer timescale. However, its application to tokamak plasma has not been fruitful due to the appearance of spurious transient oscillations in the lifting process, which are present when the kinetic simulations are initialized with a simplified model distribution function and which make the kinetic simulations to deviate from the desired paths. In this work, a kinetically informed lifting algorithm is added to the equation-free projective integration method, which is then verified in the electrostatic gyrokinetic particle-in-cell code XGCa [R. Hager and C. S. Chang, Phys. Plasmas 23, 042503 (2016)] for a neoclassical ion heat transport problem with adiabatic electrons. This new lifting operator is demonstrated to control spurious transients, enabling an over four-times reduction in the overall computing time in the time-evolution of the ion temperature profile in an axisymmetric toroidal plasma. Further reduction in the computing time is found to be limited due to the stability properties of the linear least squares projective integrator.

Transport parameters and permeation behavior of hydrogen isotopes in the first wall materials of future fusion reactors

Abstract The first wall of a future magnetic fusion device is essentially defined as the plasma-facing surface of the breeding blankets, which is supposed to be subjected to bi-directional hydrogen isotopes permeation: in one direction by edge plasma-driven permeation (PDP) of deuterium as well as tritium into blankets, and in the other direction by breed tritium gas-driven permeation (GDP) into the edge plasma. Deuterium and tritium PDP will complicate the recovery of tritium from the blanket, while tritium GDP will lead to an unwanted increase of particle recycling in the first wall region, which could even affect core confinement performance. Reduced activation ferritic/martensitic (RAFM) steels are widely proposed as candidate structural materials for the blanket of a DEMO reactor, the surface coatings made of tungsten are necessary to protect the plasma-facing wall from sputtering under high-energy particle bombardment. Therefore, the characterization of hydrogen isotopes transport through a multi-layer W + RAFM wall is of crucial importance to evaluate major reactor design issues including tritium retention, particle recycling and breeding feasibility, etc. This paper is intended to provide a review over the transport parameters of hydrogen isotopes in RAFM steels, including permeability, diffusivity, solubility and surface recombination coefficient. In addition, the present research status of hydrogen isotopes permeation and retention behavior of tungsten coated RAFM steels is briefly introduced.

Redirection of radio-frequency power flow by filaments

When radio-frequency waves propagate through magnetized plasmas, they scatter at density filaments, which are usually aligned with the magnetic field lines. We show that this scattering can redirect part of the power from the perpendicular direction to the parallel direction (along the field lines, along the filament). In magnetic confinement fusion devices, such as tokamaks, plasma is often heated by ion cyclotron range of frequency (ICRF) antennas. For such antennas, our simulations show that a parallel Poynting flux in the level of 1 MW m−2 can be generated by this power redirection mechanism. The increase of the magnitude of electric and magnetic fields of the wave in the filament is responsible for the parallel redirection of power flow. Simulations with constant density in the filament and background plasma show that the dominant parallel wavenumbers of the energy spectrum in the filament are increased and additional wavenumbers can be generated. In the cases with an experimental density profile, only the energy spectrum inside the filament is significantly affected while the rest remains almost the same.

Fiber-Optic Silicon Fabry-Perot Interferometric Bolometer: The Influence of Mechanical Vibration and Magnetic Field

Plasma radiation plays an important role in the operation and control of magnetic confinement fusion and is typically measured by bolometers. Here, we study the influence of mechanical vibration and magnetic field, which are typically present in magnetic confinement fusion devices, on the performance of a fiber-optic bolometer (FOB) based on silicon Fabry-Perot interferometer (FPI) demodulated by a scanning diode laser. It is found that both the vibration and magnetic field could cause large noise due to the birefringence of the silicon FPI. To mitigate the birefringence effects, we demonstrated two effective methods: polarization maintaining FOB and polarization scrambling. We show that two methods can be effective in mitigating the effect caused by the sensor birefringence: (1) replacing the regular single-mode fiber in the FOB system with polarization-maintaining fiber and (2) using a polarization scrambler after the laser source. For both methods, no significant increase in the noise was observed in the vibration and magnetic field tests. Our research highlights the importance of the polarization management for the FOB systems toward their practical applications in magnetic-confinement fusion systems.

An in situ diagnostic method for monitoring of fuel retention on the first wall under long-pulse operation of experimental advanced superconducting tokamak

Plasma-wall interaction (PWI) research is an active field of study in long-pulse operation in current magnetic confinement fusion devices, such as the Experimental Advanced Superconducting Tokamak (EAST). It is an urgent requirement to be able to investigate several key PWI issues, such as fuel retention, by in situ diagnostic methods. In this work, an in situ diagnostic method of laser-induced breakdown spectroscopy (LIBS) combined with optical emission spectroscopy (OES) is developed. The whole system is applied to study PWI and fill the research gap concerning the correlation between fuel retention and edge plasma conditions during long-pulse plasma operation conditions in EAST. The fuel retention intensity from LIBS on the first wall and the edge plasma condition from OES are monitored simultaneously during the long-pulse plasma operation in EAST. The results indicate that the deuterium (D) retention amount increases as the local edge D particle fluence increases. The results effectively demonstrate the potential of the LIBS method for in situ investigation of the fuel retention for PWI study in upcoming long-pulse fusion devices such as the International Thermonuclear Experimental Reactor (ITER).

Application of tungsten–copper composite heat sink materials to plasma-facing component mock-ups

The exhaust of power and particles is currently considered as one of the ultimate challenges in view of the design of a power producing magnetic confinement thermonuclear fusion device, like DEMO. One predominantly challenging aspect in this regard is the design and manufacture of divertor target plasma-facing components (PFCs) that have to sustain substantial particle, heat and neutron fluxes during fusion operation. With respect to the design of highly loaded actively cooled PFCs, copper (Cu) alloys are currently regarded as state-of-the-art structural heat sink materials (HSM). However, it has been underlined that the use of Cu alloys in PFCs implies issues mainly due to the behaviour of these materials under neutron irradiation characterised by a pronounced loss of ductility at lower and a loss of strength at elevated temperatures. These operating temperature limitations impose a strong constraint on the design of divertor PFCs and have regarding DEMO in the literature been termed a high impact design engineering risk. Against this background, the development of tungsten–copper (W–Cu) composites as potentially advanced HSMs for highly loaded PFCs was pursued by the authors during recent years. The progress of these developments is discussed in the present paper in terms of results of high-heat-flux tests conducted on PFC mock-ups that comprised W–Cu composite material heat sinks. Overall, the results of these tests indicate that W–Cu composites can indeed be regarded a viable class of advanced materials for the heat sink of highly loaded PFCs.

Influence of the inverse sheath on divertor plasma performance in tokamak edge plasma simulations

AbstractRecently, it was shown that strong electron thermionic emission from material walls could result in the formation of an “inverse sheath,” which prevents the flow of cold ions to the wall.[1–3] Such regimes look very favourably from the point of view of plasma–material interactions at the edge of magnetic fusion devices, where the problem of the erosion of plasma‐facing components under ion irradiation is one of the key issues for the development of future magnetic fusion reactors. However, it is not clear whether such regimes are compatible with edge plasma parameters and heat removal requirements in fusion reactors. To address the issue of practicality of the “inverse sheath” regime for edge tokamak plasma conditions, we perform a set of numerical simulations with 2D edge plasma transport code UEDGE[4] for a DIII‐D‐like geometry and magnetic configuration. To describe both “standard” and “inverse sheath” conditions within the framework of the UEDGE code (which does not consider the sheath region per se), at the material surfaces, we apply effective boundary conditions that emulate both “standard” and “inverse sheath” regimes. We demonstrate that, for the same input parameters, spatial distributions of edge plasma parameters corresponding to detached divertor and “inverse sheath” regimes are similar, with only a few minor differences. We discuss the compatibility of “inverse sheath” regimes with core plasma parameters.

Real-Time Control for the EHU Stellarator

At present, two main magnetic confinement fusion devices exist: tokamaks and stellarators. Moreover, stellarators have been demonstrated to be a good alternative to tokamaks, due to their ability to operate in continuous mode, which eventually translates into a higher commercial profitability. In stellarators, the magnetic confinement of the plasma is achieved exclusively by the coils, thus no electric current through the plasma is needed. In particular, this article presents the Columbia Non-Neutral Torus stellarator that is located in the Automatic Control Group of Euskal Herriko Unibertsitatea (EHU). This EHU stellarator maintains symmetry in its structure due to the topology of the mesh that is formed by its coils. A cornerstone of future fusion reactors is to obtain real-time control that enables a sustained reaction. In this article, a control-oriented model for the installed magnetic confinement coils is presented. The model is based on matrices that preserve symmetry, which is defined from physical principles and then validated by different sets of experimental data. Then, based on this model, a novel predictive control suited to this particular model with symmetric objective function is implemented in the numerical simulations, and its response is compared to that of traditional controllers. Finally, this control is implemented in a real plant and the satisfactory experiment results provide validation of both the numerical model and proposed controller.

Exploring the regime of validity of global gyrokinetic simulations with spherical tokamak plasmas

Plasma turbulence is considered one of the main mechanisms for driving anomalous thermal transport in magnetic confinement fusion devices. Based on first-principle model, gradient-driven gyrokinetic simulations have often been used to explain turbulence-driven transport in present fusion devices, and in fact, many present predictive codes are based on the assumption that turbulence is gradient-driven. However, using the electrostatic global particle-in-cell gyrokinetic tokamak simulation (GTS) code (Wang et al 2010 Phys. Plasmas 17 072511), we will show that while global gradient-driven gyrokinetic simulations provide decent agreement in ion thermal transport with a set of NBI-heated NSTX (Ono et al 2000 Nucl. Fusion 40 557) H-mode plasmas, they are not able to explain the observed electron thermal transport variation in a set of RF-heated L-mode plasmas, where a factor of 2 decrease in electron heat flux is observed after the cessation of the RF heating. Thus, identifying the regime of validity of the gradient-driven assumption is essential for first-principle gyrokinetic simulation. This understanding will help us to more confidently predict the confinement performance of ITER and future magnetic confinement devices.

Discovery of an Electron Gyroradius Scale Current Layer: Its Relevance to Magnetic Fusion Energy, Earth's Magnetosphere, and Sunspots

In the Earth’s magnetosphere, sunspots and magnetic cusp fusion devices, the boundary between the plasma and the magnetic field is marked by a diamagnetic current layer with a rapid change in plasma pressure and magnetic field strength. First principles numerical simulations were conducted to investigate this boundary layer with a spatial resolution beyond electron gyroradius while incorporating a global equilibrium structure. The boundary layer thickness is discovered to be on the order of electron gyroradius scale due to a self-consistent electric field suppressing ion gyromotion at the boundary. Formed at the scale of the electron gyroradius, the electric field plays a critical role in determining equilibrium structure and plasma transport. The discovery highlights the necessity to incorporate electron gyroradius scale physics in studies aimed at advancing our understanding of fusion devices, the magnetosphere and sunspots.

Preliminary result of cavity ring-down spectroscopy system for radio frequency negative ion source test facility.

Negative ion source is a core part of the neutral beam injection system for magnetic confinement fusion devices. The density of produced hydrogen negative ions is a critical parameter of the negative ion source. Cavity ring-down spectroscopy (CRDS) is an ultrasensitive absorption diagnostic technique for density measurement. Based on the photodetachment process, CRDS can measure the integrated line-of-sight hydrogen negative ion density in a high power ion source. The CRDS diagnostic system has been applied to Hefei utility negative ion test equipment with the radio frequency (RF) source, which is now one of the references for the China Fusion Engineering Test Reactor neutral beam injection system. Typical ring-down signals are obtained to calculate the density of hydrogen negative ions. The time evolution of hydrogen negative ion density is successfully measured. Preliminary experiments show the accurate relationship between RF power and measured hydrogen negative ion density.

Analysis of the characteristics of the plasma of an RF driven ion source for a neutral beam injector

A radio frequency (RF) driven ion source is a very important component of a neutral beam injector for large magnetic confinement fusion devices. In order to study the key technology and physics of an RF driven ion source for a neutral beam injector in China, an RF ion source test facility was developed at the Institute of Plasma Physics, Chinese Academy of Sciences. In this paper, a two-dimensional fluid model is used to simulate the fundamental physical characteristics of RF plasma discharge. Simulation results show the relationship of the characteristics of plasma (such as electron density and electron temperature) and RF power and gas pressure. In order to verify the effectiveness of the model, the characteristics of the plasma are investigated using a Langmuir probe. In this paper, experimental and simulation results are presented, and the possible reasons for the discrepancies between them are given. This paper can help us understand the characteristics of RF plasma discharge, and give a basis for further R&D for an RF ion source.

Localized divertor leakage measurements using isotopic tungsten sources during edge-localized mode-y H-mode discharges on DIII-D

Experiments carried out on DIII-D using a novel setup of isotopic tungsten (W) sources in the outer divertor have characterized how the W leakage from this region depends on both the exact source location and edge-localized mode (ELM) behavior. The sources are toroidally-symmetric and poloidally-localized to two regions: (1) the outer strike point (OSP) with natural abundance of W isotopes; and (2) the far-target with highly-enriched 182W isotopes. With the use of a dual-faced collector probe (CP) in the main scrape-off layer (SOL) near the outside midplane and source-rate spectroscopy, a proxy for divertor impurity leakage is developed. Using this proxy, it is found that for the OSP W location, there is a nearly linear increase of leakage with the power across the separatrix (), which is consistent with the effect of an increased upstream ion temperature parallel gradient force in the near-SOL; trends in the pedestal density and collisionality are also seen. Conversely, it is found that for the far-target W location leakage falls off rapidly as increases and ELM size decreases, which is suggestive that ELM size plays a role in the leakage from this location. Indications for main SOL W contamination is evidenced by the measurement of large deposition asymmetries on the two opposite CP faces. These measurements are coupled with interpretive modeling showing SOL W accumulation near the separatrix furthest from both targets driven by forces parallel to the magnetic field. This experimental setup, together with the target and upstream W measurements, provides information on the transport from different divertor W source locations and leakage. These studies help to elucidate the physics driving divertor impurity source rates and leakage, with and without ELMs, and provide better insight on the link in the chain connecting wall impurity sources to core impurity levels in magnetic fusion devices.

Multi-diagnostic analysis of plasma filaments in the island divertor

Filaments or blobs are well known structures in turbulence in magnetic fusion devices, they are considered to be the major cross-transport channel in the scrape off layer. They originate at the last closed magnetic flux surface and propagate out on the low field side of toroidal devices due to polarization in the curved magnetic field. The Wendelstein 7-X stellarator has a complex three-dimensional magnetic field structure and additionally the plasma is bounded by a chain of magnetic islands, forming an island divertor. After the first observation of filaments in Wendelstein 7-X with video cameras a multi-diagnostic study is presented in this paper to reveal their 3D structure and dynamics. Filaments are seen to be born at the edge and, at least in some cases, seen to extend to up to 4 toroidal turns. After moving radially out a few cm they enter the edge island. Here they disappear from the equatorial plane and about 200 microseconds later reappear on the outboard side of the island. A long-wavelength (∼20–30 cm) quasi coherent mode is observed in both regions where filaments appear. The similarities and differences between the filaments seen in W7-X and other devices are discussed. Possible explanations for this strange radial propagation are considered, together with the likely role of filaments in the edge and island density profile.

Doppler line shape model for the W I 5d56s—5d56p transition at 400.9 nm in tokamaks

Tungsten (W) spectroscopy is known to be a useful tool for quantifying W fluxes in tokamaks. We aim to analyze how background plasma parameters influence the W spectral-line shape measured in the divertor and scrape-off layer and what information could be retrieved from such an analysis. To that end, a Monte Carlo model is developed to simulate the line shape of the 400.9 nm W I line emitted by physically sputtered W ions from plasma facing components. The influence of background plasma parameters (ion and electron temperature) and geometrical effects (detector positioning with respect to the eroded surface) are investigated. The effect of Zeeman splitting on the spectral line-shape is also explored. A set of requirements for the experimental set-up needed to measure the analyzed spectral line features with line peak shifts of the order of 10−2Å and peak widths in the range from 10−1Å to 10−2Å are proposed. These requirements are shown to be experimentally accessible, through the installation of specific high-resolution spectrometers not typically used on magnetic fusion devices.

Nonlinear gyrokinetic Coulomb collision operator

A gyrokinetic Coulomb collision operator is derived, which is particularly useful to describe the plasma dynamics at the periphery region of magnetic confinement fusion devices. The derived operator is able to describe collisions occurring in distribution functions arbitrarily far from equilibrium with variations on spatial scales at and below the particle Larmor radius. A multipole expansion of the Rosenbluth potentials is used in order to derive the dependence of the full Coulomb collision operator on the particle gyroangle. The full Coulomb collision operator is then expressed in gyrocentre phase-space coordinates, and a closed formula for its gyroaverage in terms of the moments of the gyrocentre distribution function in a form ready to be numerically implemented is provided. Furthermore, the collision operator is projected onto a Hermite–Laguerre velocity space polynomial basis and expansions in the small electron-to-ion mass ratio are provided.