Magnetic Confinement Fusion Devices Research Articles

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.

Development of a system level code for tritium transport analysis of the WCCB blanket for CFETR

Tritium is the fuel for the future D–T magnetic-confined fusion devices such as the China Fusion Engineering Test Reactor (CFETR). Accurate assessment of the tritium transport behaviors within the reactor is essential to the fulfillment of fuel self-sustainability and public safety concerns. Due to the radioactive nature of this hydrogen isotope, numerical simulations are often used as a more convenient and realizable approach to provide data and insights regarding the evaluation of overall tritium distribution, permeation, and inventory. In this work, we developed an in-house tritium transport analysis code for the water-cooled ceramic breeder (WCCB) blanket for CFETR at the system level. The math-physics models applied incorporate the multi-physical processes required in the full functional description of macroscopic tritium transport. The code extends the capability of the existing, pioneering 0D codes in this area to take care of the varying configurations of different blanket modules within a fusion reactor. The code developed is applied as a first attempt to evaluate the overall tritium transport behaviors in the WCCB blanket system for CFETR. Preliminary simulation results are reported and discussed. It suggests that the in-homogeneity effect introduced by the geometric and material configurations varying across blanket modules can bring a marked impact on the tritium transport performance, thus is desirable to be taken into account in system-level codes to improve the prediction results.

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).

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.

Evaluation of implantation-driven tritium transport in the first wall of the WCCB blanket for CFETR

Tritium is the fuel for the future magnetic-confined fusion devices that requires accurate simulation investigation. A generic multi-physics model is implemented in this work to analyze the tritium transport behaviors in the fusion reactor components. The model is applied to evaluate plasma-induced tritium permeation and retention in the first wall (FW) of the water-cooled ceramics breeder blanket under typical operational conditions for the China Fusion Engineering Test Reactor. The total amount of tritium retention and permeation due to ion-implantation is obtained for different reactor power levels. The impact on the tritium transport behaviors from the Soret effect is also analyzed. Simulation results suggest that the tungsten armor can be a dominating source of mobile tritium retention in comparison to the steel structure in the FW; and that the Soret effect would impact the amount of tritium permeated into the coolant, in a favorable way.

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.

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.

Features of near and far scrape-off layer heat fluxes on the Wendelstein 7-X inboard limiters

The power fall-off length is used as a characteristic dimension to describe the power exhaust in the scrape-off layer (SOL) in magnetic confinement fusion devices. Measurements from the limiter campaign of the stellarator Wendelstein 7-X with inboard limiters are presented with the first-time characterization of the limiter heat loads. Two fall-off regimes are found with a narrow near SOL with power fall-off length in millimeters and a wider far SOL with fall-off lengths of several centimeters. An attempt is made to describe both regimes with a scaling law for plasma with different heating powers and densities. The results confirm the major geometry effects of the connection length on the heat transport predicted by 3D modeling.

Synthetic conventional reflectometry probing of edge and scrape-off layer plasma turbulence

Microwave reflectometry techniques have been successfully applied to measure electron density fluctuations in magnetic confined fusion devices with good spatial and time resolutions. However, quantitative interpretation of turbulence measurements has driven continuous development of both analytical theory and sophisticated numerical codes in support of reflectometry. Comparisons between experimental and synthetic reflectometry have been performed previously while only recently realistic gyro-fluid simulations have been employed together with a full-wave code to simulate measured turbulence properties with reflectometry. In this work, we report on recent efforts to employ the two-dimensional full-wave code REFMUL to implement a synthetic reflectometry diagnostic to model plasma turbulence measurements on the edge and scrape-off layer (SOL) peripheral regions of fusion devices. Numerical descriptions of microscopic turbulence were obtained from both an analytical model (following a Kolgomorov-like wavenumber spectrum) and from the GEMR turbulence code based on gyro-fluid theory. Simulations of fixed frequency conventional reflectometry with ordinary mode (O-mode) wave propagation were carried out for both turbulence cases separately. Preliminary comparisons between synthetic measurements, numerical plasma characteristics, and experimental data from ASDEX Upgrade tokamak are made. As previously described in literature, regimes of linear and non-linear response occurring at low and high turbulence levels, respectively, are observed and characterized. Synthetic spectra across moderate to high turbulence levels display qualitatively good agreement with experimental data across the edge region.

Measuring the plasma composition in tokamaks with metallic plasma-facing components

In present and future magnetic confined fusion devices with metallic plasma-facing components (PFCs) such as JET-ILW and ITER, the calculation of the plasma composition must account for multiple impurities of a wide range of mass and charge, resolve their poloidal asymmetries and account for different central peakings for various elements. Single measurements of radiation and effective charge are not enough to characterize this complex system and a self-consistent analysis of data from multiple diagnostics is required. This contribution describes a method to calculate the plasma composition simultaneously accounting for contributions of up to two low-Z impurities, and two mid-/high-Z impurities. The analysis stems from methodologies explained in Sertoli et al. (Rev. Sci. Instrum., vol. 89 (11), 2018, 113501), expanded to include more impurities and to coherently analyse multiple diagnostics within the same framework. The example Ne-seeded JET-ILW hybrid discharge reported here shows that Be, Ne, Ni and W are necessary to simultaneously explain the observed soft X-ray emission, the W concentration measured by passive vacuum ultra-violet spectroscopy, the line-of-sight integrated measurement of the effective charge, the observed poloidal asymmetry of the soft X-ray (SXR) emission, the Ne density measured by charge-exchange-recombination spectroscopy and the line-of-sight integrals of the total radiation as measured by bolometry. This consistent picture of the elemental composition enables the calculation of the radial profiles of the effective charge, the dilution and total radiation. For the cases analysed up to now, these are often very different from the typical assumptions presently used when modelling JET-ILW discharges. This will affect, among others, the calculation of neutron rates, current density profile and heat transport. These considerations are of course valid for all present and future magnetic-controlled fusion devices which exhibit multi-material plasma-facing components, including ITER.

Mineral insulated cable assessment for inductive magnetic diagnostic sensors of a hot-wall tokamak

The COMPASS-U tokamak, designed to be a 5 T magnetic field device with a full-metal first wall and operating at plasma-facing component temperatures up to 500°C, will start its operation in 2022 at IPP Prague. This device will address ITER and DEMO relevant plasma exhaust physics, including operation with liquid metal divertor. Inductive magnetic diagnostics based on conductive loops of different geometry and orientation are crucial for magnetic confinement fusion devices. Due to the high temperatures of the vacuum vessel upon which they will be operated, a suitable cable insulation needs to be chosen carefully. Mineral-insulated cables (MIC) have proven to be compatible with high baking temperatures. However, the steel sheath of MIC attenuates the response of the sensor at higher frequencies which could affect real-time plasma control feedback and magnetic equilibrium reconstruction. In this work, characterization and testing of multiple MgO MIC of different diameters was conducted. A variety of electrical property measurements, such as frequency attenuation, resistance and capacitance, for each cable is presented, both at low and high temperatures up to 300°C. Cutoff frequencies from 65 kHz to 335 kHz were identified and attributed to the shielding in a flux loop configuration. Using an external RLC circuit, the frequency response of MIC coils is compared to an electrical model for shielded coils, yielding an useful calibrated model for future probe prototypes with different geometries in the frequency range of interest.

Topology optimization of tungsten/copper structures for plasma-facing component applications

Tungsten has established itself as the most suitable plasma-facing material for long-term operation in future magnetic-confinement fusion devices, but its properties make it a poor structural material and complicate the manufacturing of complex components. Recent advances in additive-manufacturing (AM) technology have begun to make the production of tungsten components with complex geometry more feasible. The design freedom afforded by AM could be leveraged to produce more resilient plasma-facing components (PFCs). A methodology to optimize the material distribution of composite PFCs was developed to reduce the maximum thermal stress caused by high heat fluxes. Its use was demonstrated for copper-infiltrated AM tungsten (WAM/Cu) structures. Stress reductions of 50%–85% are predicted under nominal load conditions. Optimized designs also reduce stress over a wide range of off-nominal conditions. The resulting optimized structures are composed of a spatially heterogeneous distribution of W and Cu comprising a broad range of composite mixtures. A sample manufacturable component was modelled based on optimization results.Highlights• A methodology to optimize the material distribution of composite PFCs was developed to reduce the maximum thermal stress caused by high heat fluxes.• Stress reductions of up to 85% compared to a monolithic W block may be feasible with topology optimization techniques.• Optimized component designs are effective at reducing stress even over a wide range of off-nominal conditions.• Manufacturable components can be designed based on optimization results.

A scale-separated approach for studying coupled ion and electron scale turbulence

Multiple space and time scales arise in plasma turbulence in magnetic confinement fusion devices because of the smallness of the square root of the electron-to-ion mass ratio and the consequent disparity of the ion and electron thermal gyroradii and thermal speeds. Direct simulations of this turbulence that include both ion and electron space–time scales indicate that there can be significant interactions between the two scales. The extreme computational expense and complexity of these direct simulations motivates the desire for reduced treatment. By exploiting the scale-separation between ion scales (IS) and electron scales (ES), and expanding the gyrokinetic equations for the turbulence in , we derive such a reduced system of gyrokinetic equations that describes cross-scale interactions. The coupled gyrokinetic equations contain novel terms which provide candidate mechanisms for the observed cross-scale interaction. The ES turbulence experiences a modified drive due to gradients in the IS distribution function, and is advected by the IS drift, which varies in the direction parallel to the magnetic field line. The largest possible cross-scale term in the IS equations is sub-dominant in our expansion. Hence, in our model the IS turbulence evolves independently of the ES turbulence. To complete the scale-separated approach, we provide and justify a parallel boundary condition for the coupled gyrokinetic equations in axisymmetric equilibria based on the standard ‘twist-and-shift’ boundary condition. This approach allows one to simulate multi-scale turbulence using ES flux tubes nested within an IS flux tube.

Advances in experimental research towards high confinement and steady state operation on the Experimental Advanced Superconducting Tokamak

The final goal of nuclear fusion science and technology research is to develop safe and clean fusion energy which can solve the human energy crisis ultimately. As construction started in the early of this century, International Thermonuclear Experimental Reactor (ITER) project has pushed the fusion energy development a big step forward toward commercial application. Those tokamaks which are ITER-like magnetic confinement fusion devices have focused their fusion energy research on high temperature plasma experiments at this stage. The Experimental Advanced Superconducting Tokamak (EAST) was designed and constructed independently by Institute of Plasma Physics, Chinese Academy of Sciences. EAST has implemented lots of exploration and research on the key engineering and physical problems for fusion reactor. Especially in the aspect of high confinement and steady-state operation, EAST has made major experimental research progress, which is important to the commercial power generation of fusion reactors economically and practically. EAST device has achieved the 400 s long pulse plasma in low confinement mode in 2012, then in 2017 EAST refreshed the world record of high confinement mode plasma discharge operation time, and became the first tokamak device that has the capability of high confinement and steady-state operation with 100 s. Moreover, for the first time, EAST touched the longest time scale in the magnetically confined fusion plasma—particle equilibration time. In order to provide a good experimental platform for investigation of the first wall with high heat load, edge localized modes suppression and engineering and physical problems research towards the steady-state operation and control for future fusion reactor, a series of experiments were implemented on EAST, including tungsten-copper water cooled divertor, lithium first wall, current driving with radio frequency wave, projectile injection, precise magnetic surface configuration control, edge localized modes and high heat load mitigation, and a lot of valuable results have been obtained.

Characterization of wall conditions in Uragan-2M stellarator using stainless steel thermal desorption probe

The method and device have been developed and used for operative estimation of impurity level, outgassing rate and molecular layers number on high vacuum chamber surfaces of magnetic confinement fusion devices in situ. The method is based on the thermal desorption of gases into a vacuum vessel from the surface of a special metal probe during its heating up to temperature of 300 °C. To investigate and use this method, the probe was designed, manufactured and installed into the vacuum vessel of the Uragan-2 M stellarator. The technique was tested after applying VHF/RF cleaning discharges in different scenarios. A decrease of the released gas amount from the surface by more than one order of magnitude was recorded. The mass-spectrometer registered the atomic masses 18 u (H2O), 28 u (CO + N2) and 44 u (CO2), as the main gases desorbed from the probe surface during its heating. Heavy hydrocarbon masses (mainly 58 u) were also registered. The method had also been tested in the regime of hydrogen gas retained in the wall.

Design of the poloidal field system for KTX

The design of the poloidal field (PF) system includes the ohmic heating field system and the equilibrium (EQ) field system, and is the basis for the design of a magnetic confinement fusion device. A coupling between the poloidal and plasma currents, especially the eddy current in the stabilizing shell, yields design difficulties. The effects of the eddy current in the stabilizing shell on the poloidal magnetic field also cannot be ignored. A new PF system design is thus proposed. By using a low-μ material (μ = 0.001, ε = 1) instead of a conductive shell, an electromagnetic model is established that can provide a continuous eddy current distribution on the conductive shell. In this model, a 3D time-domain problem with shells translates into a 2D magnetostatic problem, and the accuracy of the calculation is improved. Based on these current distributions, we design the PF system and analyze how the EQ coils and conductive shell affect the plasma EQ when the plasma ramps up. To meet the mainframe design requirements and achieve an efficient power-supply design, the position and connection of the poloidal coils are optimized further.

Time-domain global similarity method for automatic data cleaning for multi-channel measurement systems in magnetic confinement fusion devices

To guarantee the availability and reliability of data source in Magnetic Confinement Fusion (MCF) devices, incorrect diagnostic data, which cannot reflect real physical properties of measured objects, should be sorted out before further analysis and study. Traditional data sorting cannot meet the growing demand of MCF research because of the low-efficiency, time-delay, and lack of objective criteria. In this paper, a Time-Domain Global Similarity (TDGS) method based on machine learning technologies is proposed for the automatic data cleaning of MCF devices. The aim of traditional data sorting is to classify original diagnostic data sequences. The lengths and evolution properties of the data sequences vary shot by shot. Hence the classification criteria are affected by many discharge parameters and are different in various discharges. The focus of the TDGS method is turned to the physical similarity between data sequences from different channels, which are more independent of discharge parameters. The complexity arisen from real discharge parameters during data cleaning is avoided in the TDGS method by transforming the general data sorting problem into a binary classification problem about the physical similarity between data sequences. As a demonstration of its application to multi-channel measurement systems, the TDGS method is applied to the EAST POlarimeter–INTerferometer (POINT) system. The optimal performance of the method evaluated by 24-fold cross-validation has reached 0.9871 ± 0.0385.