Steady-state Tokamak Operation Research Articles

The influence of E× B drift on tungsten target erosion and W impurity transport during neon seeding on EAST

Mitigating tungsten (W) wall erosion and core accumulation are vitally important for the steady-state operation of tokamaks. It is well known that drifts have a great impact on the transport of charged particles in the edge region, which could affect W source and W impurity transport. In this work, SOLPS-ITER modeling is applied to study the W impurity behavior on EAST during neon seeding with the consideration of E× B drift. The objective is to establish the relationship between the eroded W flux, W transport and the corresponding accumulation in the core in different discharge regimes. The effects of drift on W sputtering at targets and W impurity distribution in the cases of different toroidal magnetic field (B t) directions are assessed. The simulation results indicate that drift could influence W transport via W impurity retention and redistribution in the divertor, and the leakage from the divertor. In forward B t (B× ∇B points to the X-point), eroded W flux at the outer target is increased remarkably, and most of the W ions transport from the outer to the inner divertor and escape to the upstream region in the high field side. W ions mainly transport from the inner to the outer divertor and escape from the divertor in the low field side in reversed B t due to the opposite drift flux. The Ne puffing rate is scanned in forward B t and without-drift cases to further investigate the W erosion and W impurity transport in different divertor regimes. It is found that the W source from targets is generally enhanced by drift compared to cases without drift. The core accumulation, as well as poloidal asymmetry, is also influenced significantly by the drift. In the attached regime, the intense W source and strong drift flux lead to enhanced W accumulation in the core, and obvious poloidal asymmetry of W density distribution appears. The drift flux is reduced and W erosion is suppressed after detachment. W concentration in the core and poloidal asymmetry consequently decline. Therefore, adequate Ne impurity seeding can be applied to control the W accumulation in the core.

RF conditioning to suppress multipactor discharge for helicon wave current drive in KSTAR

The efficient off-axis non-inductive current drive using auxiliary power is essential for the steady state operation of tokamaks. In previous studies, a helicon wave current drive has been proposed for high beta plasma in several fusion devices, and the 476-MHz radio-frequency (RF) system for the proposed helicon wave has been installed in KSTAR. However, multipactor discharge occurred inside the waveguide system, including the vacuum feedthrough (VFT), and it interrupted the normal RF propagation to plasma. Thus, for successful helicon wave current drive, suppressing the multipactor discharge in the waveguide system, including the 6–1/8″ coaxial type VFT and the newly manufactured helical type traveling wave antenna (TWA) is essential before installation in KSTAR. RF conditioning utilizing a short pulse modulation during a long period helps suppress the multipactor discharge, enabling the 1-MW 476-MHz RF to pass through the entire waveguide system despite the existence of multipactor discharge. Following RF conditioning in VFT with a short pulse modulation, the newly manufactured helical type TWA and VFT for helicon wave current drive have been successfully installed in KSTAR.

Application of integrated simulation environment SIEMNED to the analysis of the MEPHIST-0 tokamak operation

This paper presents the results of numerical simulation of plasma equilibrium and stability in the MEPHIST-0 tokamak with SIEMNED software and comparison of simulation results with experiments. The determined characteristics of the vacuum chamber show that it significantly affects the entire discharge. For various scenarios of the inductor operation, a comparison of experimental data and simulated currents and magnetic fields induced in the chamber was carried out. For steady-state tokamak operation, a numerical study of equilibrium plasma configurations was carried out depending on the currents in the poloidal magnetic field coils and plasma current. The vertical plasma instability was investigated. The limiting values of plasma ellipticity preventing the vertical plasma instability were numerically determined. Numerical simulations show that plasma equilibrium is supported by induced currents. It was shown numerically that magnetic configuration with ‘zero of higher order’ were obtained before the plasma shot, suggesting consistency between the simulation results and observations.

Investigation of particle exhaust from EAST divertor

Particle control is one of the key issues for steady-state tokamak operation. The density decay time is used to evaluate the capability of the divertor to exhaust particle with all fueling turn off. Experiments of the EAST divertor's ability to exhaust helium particles were also performed during the EAST helium plasma operation. The experimental results show that the EAST divertor is less capable of exhaust helium particles than deuterium particles. And similar to deuterium, the helium particle exhaust performance is sensitivity on the different strike point location. During 2021 experiment campaign, a new lower tungsten divertor has been developed and installed on EAST, the particle exhaust capacity of the new divertor has also been investigated during this campaign.

Three-dimensional modeling of sputtered materials transport in diagnostic ducts of fusion devices

Migration of plasma erosion products in plasma facilities is studied experimentally and numerically within the framework of modeling transport of plasma-facing materials in the diagnostic ducts of fusion devices. Material transport simulation is discussed for two cases of low and high background neutral gas pressures. Monte Carlo software KITe was used to simulate transport at a neutral gas background pressure 0.1–0.5 Pa—typical during steady-state tokamak operation and during pressure pulses caused by edge localized modes (ELMs). The simulation approach was implemented to describe experiments at the MAGNUM-PSI facility. Fluid dynamic code FLUENT is used to simulate transport during pressure surges as high as 1000 Pa, which can occur in the case of severe disruptions in tokamak plasma discharges, such as vertical displacement events (VDE) or accidental events. The hydrodynamic approach was verified in simulation of target sputtering in the QSPA plasma gun facility.

Progress of Divertor Heat and Particle Flux Control in EAST for Advanced Steady-State Operation in the Last 10 Years

Active control of the excessively high heat and particle fluxes on the divertor target plates is of fundamental importance to the steady state operation of tokamaks, especially for fusion reactors. A series of experiments have been carried out on this critical issue to relieve the plasma-wall interactions in the experimental advanced superconducting tokamak (EAST) in the last ten years, not only contributing to the long pulse operation of EAST itself, but also providing physical understandings and potential techniques to the next-generation devices like ITER. We have characterized the power deposition pattern and broadened the divertor footprint width effectively. The plasma-wetted area is actively handled using either 3-dimentional edge magnetic topology or advanced plasma equilibrium, thereby peak heat flux around the strike point is reduced. Active control of detachment or radiation compatible with core plasma performance has progressed significantly in very recent years, with a series of active feedback control modules developed and utilized successfully, based on the divertor physics advances with both experiments and simulation. The upper divertor of EAST was upgraded from graphite to active water-cooling ITER-like tungsten in 2014, exhibiting much enhanced heat removal capability. As for the particle exhaust including both fueling and impurity particles, in addition to wall conditioning and impurity source control, the efficiency of particle flux exhaust is optimized by making full use of the divertor closure and the plasma drifts in both scrape-off layer and divertor volume. These heat and particle exhaust advances contribute greatly to a series of EAST achievements like H-mode operation over 100 s. A brief near-term plan on the integrated control of divertor plasma-wall interactions in long-time scale will also be introduced, aiming to provide favorable divertor operation solution for ITER and CFETR.

An intelligent controller design based on the neuroendocrine algorithm for the plasma density control system on Tokamak devices

Abstract Realization of real-time and accurate control of plasma electron density is one of the key points for the long-time steady-state operation of Tokamak. Based on the mechanism model of the neuroendocrine regulatory principle, this paper designs an intelligent controller and studies its application on the plasma density control system (PDCS) for the J-TEXT Tokamak. This PDCS uses the HCN laser interferometer to obtain the density signal, and controls the gas puffing valve through the proposed intelligent control algorithm to achieve accurate density control. The controller mainly includes a hypothalamic regulation module, a single neuron proportional-integral-differential (PID) control module and an ultrashort feedback module. It is designed and referenced to the long feedback, short feedback, and ultra-short feedback loop mechanisms of the human neuroendocrine hormone regulation. The proposed controller makes the PID parameters be adjusted on-line in real-time and achieves rapid and stable elimination of errors. The simulations and experimental results show that the proposed method, compared with the conventional PID control method, can synthetically improve the performance of the control system.

High voltage electrical system of 8.56 GHz CW klystron for electron cyclotron heating on QUEST spherical tokamak

A high voltage DC power supply for the cathode of the 8.56 GHz CW klystron has been set up for electron cyclotron heating (ECH) in steady state tokamak operation on QUEST spherical tokamak. The power supply is equipped with an IGBT array and a reactor for fast shutoff of the voltage in 10 μs, where the influx of electric energy at the short circuit is limited to 5 J. AC switches also have been installed in the three-phase power lines. High voltage relays are useful to save electric energy consumption. Fast three-phase AC switching by IGBT-stack is applicable to reduce the electric load of the components of the klystron power supply.

Observation of Efficient Lower Hybrid Current Drive at High Density in Diverted Plasmas on the Alcator C-Mod Tokamak.

Efficient lower hybrid current drive (LHCD) is demonstrated at densities up to n[over ¯]_{e}≈1.5×10^{20} m^{-3} in diverted plasmas on the Alcator C-Mod tokamak by operating at increased plasma current-and therefore reduced Greenwald density fraction. This density exceeds the nominal "LH density limit" at n[over ¯]_{e}≈1.0×10^{20} m^{-3} reported previously, above which an anomalous loss of current drive efficiency was observed. The recovery of current drive efficiency to a level consistent with engineering scalings is correlated with a reduction in density shoulders and turbulence levels in the far scrape-off layer. Concurrently, rf wave interaction with the edge and/or scrape-off-layer plasma is reduced, as indicated by a minimal broadening of the wave frequency spectrum measured at the plasma edge. These results have important implications for sustaining steady-state tokamak operation and indicate a pathway forward for implementing efficient LHCD in a reactor.

Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN, despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.

Neutron emission in neutral beam heated KSTAR plasmas and its application to neutron radiography

The main mission of Korea Superconducting Tokamak Advanced Research (KSTAR) program is exploring the physics and technologies of high performance steady state Tokamak operation that are essential for ITER and fusion reactor. Since the successful first operation in 2008, the plasma performance is enhanced and duration of H-mode is extended to around 50s which corresponds to a few times of current diffusion time and surpassing the current conventional Tokamak operation. In addition to long-pulse operation, the operational boundary of the H-mode discharge is further extended over MHD no-wall limit(βN∼4) transiently and higher stored energy region is obtained by increased total heating power (∼6MW) and plasma current (Ip up to 1MA for ∼10s). Heating system consists of various mixtures (NB, ECH, LHCD, ICRF) but the major horse heating resource is the neutral beam(NB) of 100keV with 4.5MW and most of experiments are conducted with NB. So there is a lot of production of fast neutrons coming from via D(d,n)3He reaction and it is found that most of neutrons are coming from deuterium beam plasma interaction. Nominal neutron yield and the area of beam port is about 1013–1014/s and 1m2 at the closest access position of the sample respectively and neutron emission could be modulated for application to the neutron radiography by varying NB power. This work reports on the results of quantitative analysis of neutron emission measurements and results are discussed in terms of beam-plasma interaction and plasma confinement. It also includes the feasibility study of neutron radiography using KSTAR.

Modeling of high harmonic fast wave current drive on EAST tokamak

High harmonic fast waves (HHFW) are among the candidates for non-inductive current drive (CD), which is essential for long-pulse or steady-state operation of tokamaks. Current driven with HHFW in EAST tokamak plasmas is numerically studied. The HHFW CD efficiency is found to increase non-monotonically with the wave frequency, and this phenomenon is attributed to the multi-pass absorption of HHFW. The sensitivity of CD efficiency to the value of the parallel refraction index of the launched wave is confirmed. The quasilinear effects, assessed as significant in HHFW current drive with the GENRAY/CQL3D package, cause a significant increase in CD efficiency as RF power is increased, which is very different from helicon current drive. Simulations for a range of toroidal dc electric fields, in combination with a range of fast wave powers, are also presented and indicate that the presence of the DC field can also enhance the CD efficiency.

High internal inductance for steady-state operation in ITER and a reactor

Increased confinement and ideal stability limits at relatively high values of the internal inductance () have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. Normalized plasma pressure in the range appropriate for a reactor has been achieved in high elongation and triangularity double-null divertor discharges with at , near the ideal kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at . Confinement is above the H-mode level with . At , the current is overdriven, with bootstrap current fraction , noninductive current fraction and negative surface voltage. For ITER (which has a single-null divertor shape), operation at is a promising option with and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at , so far reaching and at with performance appropriate for the ITER Q=5 mission, . Modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high discharge. Stable solutions in the double-null shape are found without the vacuum vessel wall at , and , and at with the vacuum vessel wall.

Global gas balance and influence of atomic hydrogen irradiation on the wall inventory in steady-state operation of QUEST tokamak

Hydrogen wall pumping is studied in steady state tokamak operation (SSTO) of QUEST with all metal plasma facing materials PFMs at 100°C. The duration of SSTO is up to 820s in fully non-inductive plasma. Global gas balance analysis shows that wall pumping at the apparent (retention–release) rate of 1–6×1018H/s is dominant and 70–80% of injected H2 can be retained in PFMs. However, immediately after plasma termination the H2 release rate enhances to ∼1019H/s. In order to understand a true retention process the direct measurement of retention flux has been carried out by permeation probes. The comparison between the evaluated wall retention and results from global analysis is discussed.

Advances in multi-megawatt lower hybrid technology in support of steady-state tokamak operation

It has been demonstrated that lower hybrid current drive (LHCD) systems play a crucial role for steady-state tokamak operation, owing to their high current drive (CD) efficiency and hence their capability to reduce flux consumption. This paper describes the extensive technology programmes developed for the Tore Supra (France) and the KSTAR (Korea) tokamaks in order to bring continuous wave (CW) LHCD systems into operation. The Tore Supra LHCD generator at 3.7 GHz is fully CW compatible, with RF power PRF = 9.2 MW available at the generator to feed two actively water-cooled launchers. On Tore Supra, the most recent and novel passive active multijunction (PAM) launcher has sustained 2.7 MW (corresponding to its design value of 25 MW m−2 at the launcher mouth) for a 78 s flat-top discharge, with low reflected power even at large plasma-launcher gaps. The fully active multijunction (FAM) launcher has reached 3.8 MW of coupled power (24 MW m−2 at the launcher mouth) with the new TH2103C klystrons. By combining both the PAM and FAM launchers, 950 MJ of energy, using 5.2 MW of LHCD and 1 MW of ICRH (ion cyclotron resonance heating), was injected for 160 s in 2011. The 3.7 GHz CW LHCD system will be a key element within the W (for tungsten) environment in steady-state Tokamak (WEST) project, where the aim is to test ITER technologies for high heat flux components in relevant heat flux density and particle fluence conditions. On KSTAR, a 2 MW LHCD system operating at 5 GHz is under development. Recently the 5 GHz prototype klystron has reached 500 kW/600 s on a matched load, and studies are ongoing to design a PAM launcher. In addition to the studies of technology, a combination of ray-tracing and Fokker–Planck calculations have been performed to evaluate the driven current and the power deposition due to LH waves, and to optimize the N∥ spectrum for the future launcher design. Furthermore, an LHCD system at 5 GHz is being considered for a future upgrade of the ITER Heating and Current Drive systems, with a power capability of 20 MW coupled to the plasma using a PAM launcher. An R&D programme is being conducted at CEA/IRFM to develop a BeO vacuum window which is a safety critical component of the transmission line. In addition, a mock-up of a TE10–TE30 mode converter at 5 GHz, designed for a rectangular transmission line, has been manufactured and successfully tested on Tore Supra at low RF power.

A quasilinear formulation of turbulence driven current

Non-inductive current drive mechanisms, such as the familiar neoclassical bootstrap current correspond to an essential component to the realization of steady state tokamak operation. In this work, we discuss a novel collisionless mechanism through which a mean plasma current may be driven in the presence of microturbulence. In analogy with the traditional neoclassical bootstrap current drive mechanism, in which the collisional equilibrium established between trapped and passing electrons results in the formation of a steady state plasma current, here we show that resonant scattering of electrons by drift wave microturbulence provides an additional means of determining the equilibrium between trapped and passing electrons. The resulting collisionless equilibrium is shown to result in the formation of an equilibrium current whose magnitude is a function of the thermodynamic forces. A mean field formulation is utilized to incorporate the above components into a unified framework through which both collisional as well as collisionless current drive mechanisms may be self-consistently treated. Utilizing a linearized Fokker-Planck collision operator, the plasma current in the presence of both collisions as well as turbulent stresses is computed, allowing for the relative strength of these two mechanisms to be quantified as a function of collisionality and fluctuation amplitude.

Simulation of Lower Hybrid Current Drive for DEMO

Lower hybrid (LH) wave is not only convenient to generate a flat or reversed magnetic shear profiles, but also helps one to explore scenarios for steady-state tokamak operation with improved confinement. Here with LSC code (lower hybrid simulation code), we calculate density and temperature profiles, relative power of injected wave and current wave lunch for two options of DEMO at the launched LH wave frequency 5 GHz. Two plasma scenarios pertaining to two different DEMO options, known as pulsed (option 1) and steady-state (option 2) models, have been analyzed. We perceive that power deposition by using lower hybrid wave injection mainly takes place near the edge of plasma and approximately in more peripheral region for both of options but has approximately higher efficiency for option 1 compared to option 2. About current wave lunch, a major part of that is close to the plasma edge for both of options. We have some considerable parts that reach to internal layers for option 1 and then current drive mainly takes place in a wider, more peripheral region for option 1.

A variational multi-symplectic particle-in-cell algorithm with smoothing functions for the Vlasov-Maxwell system

Smoothing functions are commonly used to reduce numerical noise arising from coarse sampling of particles in particle-in-cell (PIC) plasma simulations. When applying smoothing functions to symplectic algorithms, the conservation of symplectic structure should be guaranteed to preserve good conservation properties. In this paper, we show how to construct a variational multi-symplectic PIC algorithm with smoothing functions for the Vlasov-Maxwell system. The conservation of the multi-symplectic structure and the reduction of numerical noise make this algorithm specifically suitable for simulating long-term dynamics of plasmas, such as those in the steady-state operation or long-pulse discharge of a super-conducting tokamak. The algorithm has been implemented in a 6D large scale PIC code. Numerical examples are given to demonstrate the good conservation properties of the multi-symplectic algorithm and the reduction of the noise due to the application of smoothing function.

Assessment of DEMO challenges in technology and physics

The challenges that DEMO designs encounter in both technology and physics are reviewed. It is shown that it is very important to respect the interlinks between these fields when developing designs for DEMO. Examples for areas where such interlinks put very strict requirements are the development of a steady state tokamak operation scenario and the question of power exhaust taking into account the boundary conditions set by materials questions. Concerning steady state operation, we find that demands on the physics scenario are so high that pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. Alternatively, the device could foresee a large fraction of externally driven current which calls for optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system. In the exhaust area, a realistic estimate of the admissable time averaged peak heat flux at the target is of the order of 5MW/m2, leading to strict requirements for the operational scenario, which has to rely on an unprecedented high level of radiation loss by impurity seeding and the facilitation of partial detachment. Thus, exhaust scenarios along these lines have to be developed which are compatible with the confinement needs and the H-L back transition power for DEMO. In both areas, we discuss possible risk mitigation strategies based on conceptually different approaches.

Recent lithium experiments in tokamak T-11M

The T-11M lithium program is focused to find solutions for technological issues of a steady-state tokamak. A recent lithium activity in T-11M was focused on three directions: investigation of different Li-limiters, investigations of long-term effects of lithium behavior in hydrogen plasmas and development of a new kind of CPS limiters. The new so-called R limiter has been used as Li collector for investigation of lithium fluxes near the plasma boundary and chamber walls. It was shown, that the Li circulation near the limiter exceeds up to four times the primary lithium flux from limiters. 90% of primary flux is collected by lateral sides of limiter and only 10% arrives at the chamber wall. The deuterium glow discharge was used to test the long-term lithium degradation under the deuterium bombardment. It was shown that such degradation is small. That means that the Li-limiter can be used as Li-emitter in steady-state tokamak operation. A new vertical lithium limiter was successfully tested in T-11M.