High Density Tokamak Research Articles

Fusion pilot plant performance and the role of a sustained high power density tokamak

Recent U.S. fusion development strategy reports all recommend that the U.S. should pursue innovative science and technology to enable construction of a fusion pilot plant (FPP) that produces net electricity from fusion at low capital cost. Compact tokamaks have been proposed as a means of potentially reducing the capital cost of a FPP. However, compact steady-state tokamak FPPs face the challenge of integrating a high fraction of self-driven current with high core confinement, plasma pressure, and high divertor parallel heat flux. This integration is sufficiently challenging that a dedicated sustained-high-power-density (SHPD) tokamak facility is proposed by the U.S. community as the optimal way to close this integration gap. Performance projections for the steady-state tokamak FPP regime are presented and a preliminary SHPD device with substantial flexibility in lower aspect ratio (A = 2–2.5), shaping, and divertor configuration to narrow gaps to an FPP is described.

Concept Design and Analysis of the Magnet System of the Sustained High Power Density Tokamak Facility

A recent National Academy study recommended a next-step sustained high power density (SHPD) facility for the United States that can be a bridge to a compact fusion pilot plant. A design has been initiated to investigate a possible SHPD non-deuterium-tritium device that builds upon recent low aspect ratio tokamak studies. A 1.2-m device with a 2.4 aspect ratio has been chosen to evaluate physics performance goals within a double-null plasma machine arrangement centered on three component features: high current density toroidal field (TF) and central ohmic heating/poloidal field (PF) coils, liquid metal divertor/first wall systems, and the integration of a limited set of outboard dual-coolant lead lithium test blankets that can be maintained within a vertical maintenance approach. The underlying initiative of the U.S. program is to promote research and technology advancements that lead to the construction of a compact pilot plant that produces electricity from fusion at the lowest possible capital cost. High-field, higher power density will bring down the size of a fusion device and can lower the cost, but this can only occur with the development of high-current density TF and PF windings, a support system to handle the higher magnetic loads, and a cost structure that is economically viable. High-temperature superconductors (HTSs) offer the potential to meet high-field/high-current requirements and is under development by a number of institutions. Fabrication process improvement would continue to bring down the price if a market for fusion devices were to develop and the production level of HTS cable were to increase; however, this is not happening with any expedience especially related to fusion applications. At present, HTS conductors are an order of magnitude more expensive than low-temperature superconductors (LTSs), making an early application for a near-term SHPD experimental device problematic, especially for a low aspect ratio design that has minimum component space within the device inboard region. In lieu of an HTS-based, high-current density TF winding design, a moderately high-current density winding design (>80 MA/m2) based on cable-in-conduit LTS conductors in a multiwinding pack arrangement sized for low and high field was defined. This paper focuses on the design of the magnet system, with attention given to the analysis performed on the SHPD TF casing structure. The analysis is carried out for various sections: (1) Emag analysis of the SHPD structure, (2) TF casing winding pack evaluation for TF-TF self-load only with slit model and three-dimensional geometry, (3) effect of the shear pin in the TF casing structure, and (4) compression ring effect on the SHPD structure.

Could Tokamaks Be the Vaccine for the Climate Change Pandemic?

The replacement of the burning of fossil fuels in power plants with other forms of clean energy, for example, that of a tokamak fusion reactor employing the deuterium-tritium cycle, like ITER, would contribute enormously to the mitigation of climate change. Unfortunately, for such a type of fusion reactor, we expect the neutrons, which carry 80% of the fusion power with energies seven times larger than those of neutrons of fission reactors, to cause serious radiation damage with possible fracture of the blanket modules and the reactor wall. Hence, before contemplating the use of tokamaks for replacing fossil fuels of conventional power plants, we need a thorough investigation of the damage caused by neutrons in high-power tokamak reactors. Unfortunately, ITER will not provide any exhaustive information since it is neither a high power density tokamak nor a reactor. However, a rise in toroidal magnetic field by a factor of 2 would bring the fusion power of ITER to 8 GW and allow an investigation of the damage caused by neutrons to internal components and the reactor wall.

Scrape-off layer transport and filament characteristics in high-density tokamak regimes

A detailed cross-device investigation on the role of filamentary dynamics in high-density regimes has been performed within the EUROfusion framework, comparing the ASDEX Upgrade (AUG) and TCV tokamaks. Both devices run density ramp experiments at different levels of plasma current, keeping the toroidal field or q95 constant in order to disentangle the role of the parallel connection length and the current. During the scan at a constant toroidal field, in both devices the scrape-off layer (SOL) profiles tend to develop a clear SOL density shoulder at a lower edge density whenever the current is reduced. Different current behaviour is substantially reconciled in terms of the edge density normalized to the Greenwald fraction. During the scan at constant q95 AUG exhibits similar behaviour, whereas in TCV no upstream profile modification signature has been observed at lower current levels. The latter behaviour has been ascribed to the lack of target density rollover. The relation between the upstream density profile modification and detachment condition has been investigated. For both devices the relation between blob size and the SOL density e-folding length is found independent of the plasma current, with the observation of a clear increase in blob size and the edge density normalized to a Greenwald fraction. ASDEX Upgrade has also explored filamentary behaviour in the H-mode. The experiments in AUG have focused on the role of neutrals, performing discharges with and without cryogenic pumps, highlighting how high neutral pressure, not only in the divertor but also at the midplane, is needed in order to develop an H-mode SOL profile shoulder in AUG.

Development of Solenoid-Driven and Pneumatic Punches for Launching High-Z Cryogenic Pellets for Tokamak Disruption Mitigation Experiments

Mitigation of disruption events in future high energy density tokamaks is essential for machine longevity. The creation of runaway electrons, large electromagnetic forces, and high localized heat loads during a disruption can be devastating to machine components. Shattered pellet injection is currently the most effective method of disruption mitigation. Injection of cryogenically solidified deuterium, neon, or argon (or mixtures thereof) have been shown to efficiently radiate thermal energy of the plasma so that the heat load is distributed on the walls of the machine. Pellets are formed by desublimating gas in the barrel of a pipe gun and fired using a pulse of high-pressure light gas. Current gas gun designs cannot reach sufficient pressure to dislodge pure neon and argon pellets at low temperatures because the material strength is too high. Pellet temperatures must be kept low (to well below the triple-point temperature of the material) to ensure minimal gas flow into the machine due to vapor pressure of the pellet. A gas-driven punch device has been designed and tested to dislodge pure neon or argon pellets. The breakaway strength of a pellet is proportional to the surface area of the pellet in contact with the inner diameter of the barrel. As pellets get larger in diameter, the amount of force needed to dislodge them increases. To better understand the mechanics behind how a punch dislodges a pellet, a solenoid-operated punch was designed so that kinetic energy of the punch, when striking a pellet, can be varied by changing input current to the solenoid. This solenoid punch will be used to determine kinetic energy versus pellet surface area threshold for breakaway. These data will be used to design mechanical punches for use in a high-field tokamak environment. This paper outlines the modeling, design, experimental testing, and results of the punch development activities.

Modelling of LHCD in high magnetic field, high density tokamak configuration.

New modelling results (performed by means of “Raystar” code, a 2D ray-tracing-Fokker-Planck numerical tool) are presented with the aim of assessing the radial LH-driven current profiles in tokamak devices characterized by high density and high magnetic field as recently proposed by MIT. The modelling calculation is performed as function of plasma and antenna parameters, in particular the peak n//, the width of the launched RF power spectrum, the coupled power, the launched position of the antenna (high field side injection) and the kinetic profiles. Moreover by launching LH antenna spectra with a suitable n// width the power spectrum propagating in the plasma is weakly modified by quasi-linear broadening effects. Non-linear interactions with plasma density fluctuations of the thermal background are also negligible, as well as the PI at the plasma edge. The LHCD efficiency has been calculated and correlated to these parameters.

Developing physics basis for the snowflake divertor in the DIII-D tokamak

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power –5 MW and a range of plasma currents MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10–15% of total outer divertor power. Measurements of electron pressure and poloidal beta support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4–4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies conducted in high-power H-mode discharges in the NSTX and DIII-D tokamaks, and, along with snowflake divertor results from TCV and other tokamaks, contribute to the physics basis of the SF divertor as a power exhaust concept for future high power density tokamaks.

Measurements of the parallel wavenumber of lower hybrid waves in the scrape-off layer of a high-density tokamak

In lower hybrid current drive (LHCD) experiments on tokamaks, the parallel wavenumber of lower hybrid waves is an important physics parameter that governs the wave propagation and absorption physics. However, this parameter has not been experimentally well-characterized in the present-day high density tokamaks, despite the advances in the wave physics modeling. In this paper, we present the first measurement of the dominant parallel wavenumber of lower hybrid waves in the scrape-off layer (SOL) of the Alcator C-Mod tokamak with an array of magnetic loop probes. The electric field strength measured with the probe in typical C-Mod plasmas is about one-fifth of that of the electric field at the mouth of the grill antenna. The amplitude and phase responses of the measured signals on the applied power spectrum are consistent with the expected wave energy propagation. At higher density, the observed k|| increases for the fixed launched k||, and the wave amplitude decreases rapidly. This decrease is correlated with the loss of LHCD efficiency at high density, suggesting the presence of loss mechanisms. Evidence of the spectral broadening mechanisms is observed in the frequency spectra. However, no clear modifications in the dominant k|| are observed in the spectrally broadened wave components, as compared to the measured k|| at the applied frequency. It could be due to (1) the probe being in the SOL and (2) the limited k|| resolution of the diagnostic. Future experiments are planned to investigate the roles of the observed spectral broadening mechanisms on the LH density limit problem in the strong single pass damping regime.

ADX: a high field, high power density, advanced divertor and RF tokamak

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ∼ 1.5 MW m−2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency 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 favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv:1409.3540)) that makes use of high-temperature superconductor technology—a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

Absorption of lower hybrid waves in the scrape off layer of a diverted tokamak

The goal of the Lower Hybrid Current Drive (LHCD) system on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] is to investigate current profile control under plasma conditions relevant to future tokamak experiments. Experimental observations of a LHCD “density limit” for C-Mod are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops suddenly above line averaged densities of 1020 m−3 (ω/ωLH∼3–4), well below the density limit previously observed on other experiments (ω/ωLH∼2). Electric currents flowing through the scrape off layer (SOL) between the inner and outer divertors increase dramatically across the same density range that the core bremsstrahlung emission drops precipitously. These experimental x-ray data are compared to both conventional modeling, which gives poor agreement with experiment above the density limit and a model including collisional absorption in the SOL, which dramatically improves agreement with experiment above the observed density limit. These results show that strong absorption of LH waves in the SOL is possible on a high density tokamak and the SOL must be included in simulations of LHCD at high density.

The super X divertor (SXD) and a compact fusion neutron source (CFNS)

A new magnetic geometry, the super X divertor (SXD), is invented to solve severe heat exhaust problems in high power density fusion plasmas. SXD divertor plates are moved to the largest major radii inside the TF coils, increasing the wetted area by 2–3 and the line length by 2–5. Two-dimensional fluid simulations with SOLPS (Schneider et al 2006 SOLPS 2-D edge calculation code Contrib. Plasma Phys. 46) show a several-fold decrease in divertor heat flux and plasma temperature at the plate. A small high power density tokamak using SXD is proposed, for either (1) useful fusion applications using conservative physics, such as a component test facility (CTF) or fission–fusion hybrid, or (2) to develop more advanced physics modes for a pure fusion reactor in an integrated fusion environment.

KSTAR 토카막 기본운전을 위한 연료주입 모의실험

KSTAR 토카막은 현재 주장치 조립을 완료하고 "최초 플라즈마" 발생과 시운전을 준비하고 있다. 이를 위해 설치되어 있는 연료주입계는 배기 덕트 반대 쪽 한 곳만 주입구로 사용하기 때문에 균일하고 신속한 입자공급이 어렵다. 시운전이 끝나고 기본운전 단계가 되면 단순한 플라즈마 생성이 아니라 일정한 고밀도 토카막플라즈마 상태를 일정시간 고르게 유지해야 하므로 공간적으로 균일한 입자공급과 플라즈마 밀도 변화에 따른 신속한 제어가 가능한 연료공급계로 개선되어야 한다. 이런 개선점을 찾는 작업의 일환으로 우선 D-형 진공용기 및 플라즈마 컬럼을 사각단면 토러스로 묘사하는 모델을 만들고 연료공급 위치와 플라즈마의 전리율 및 배기확률에 따라 공간적인 입자분포가 어떻게 바뀌는지 몬테카를로 계산을 통해 분석했다. The assembly of the main system of the KSTAR tokamak has been recently completed, and the preparation for the 1st plasma and test operations is progressed. The fueling system established for these purposes uses only one port placed at the opposite side of the pumping duct, and has a difficulty of attaining a uniform and fast supply of fuel particles to the plasma. At the base operation stage after finishing the test operation, the fueling system must be improved to provide a uniform fueling and a feed-back control in accordance with a high-density tokamak plasma maintained for a long period. As a part for understanding the points to be improved in the fueling system, a Monte Carlo simulation on the gas fueling into the tokamak plasma has been executed. After modeling the vacuum vessel and the plasma of quasi-D shapes as tori of rectangular cross-sections, the influences of the position and the number of the fueling inputs on the particle density distribution for a given pumping probability and mean free path were investigated.

Electron Bernstein Wave Heating and Emission in the TCV Tokamak

Electron cyclotron resonance heating (ECRH) of high-density tokamak plasmas is limited because of reflections of the waves at so-called wave cutoffs. Electron Bernstein wave (EBW) heating (EBWH) via a double mode conversion process from ordinary (O)-mode, launched from the low field side, to extraordinary (X)-mode and finally to Bernstein (B)-mode offers the possibility of overcoming these density limits.In this paper, the O-X mode conversion dependence on the microwave injection angle is demonstrated experimentally. The dependence on the injection angle is studied in high-density plasmas in H-mode, in the presence of magnetohydrodynamic activity, edge-localized modes, and sawteeth. The results of localized heat deposition at an overdense location are presented, demonstrating EBWH for the first time via the O-X-B mode conversion process in a standard aspect-ratio tokamak. The results of global and local power deposition are compared with ray-tracing calculations. Moreover, a temperature increase due to EBWH is observed.Initial EBW emission measurements with a newly installed ECRH reception launcher are presented. The inverse double mode conversion process B-X-O is observed by measuring the emission for several frequencies at an optimum angle.

Effect of the magnetohydrodynamic modes on generation of the nonthermal electrons in high density tokamak plasma

Various types of nonthermal x-ray bursts are observed in the central part of plasma with high density in the T-10 tokamak [Alikaev et al., Plasma Phys. Controlled Fusion 30, 381 (1988)] with the use of a tangential view x-ray imaging system. In contrast with the “classical” hard x-ray bursts produced during the interaction of the runaway electron beams (Eγ∼0.5–3 MeV) with the tokamak wall, the “internal” bursts are associated with forward bremsstrahlung emission of the suprathermal electrons (Eγ∼25–100 keV) from the plasma area around the q=1 and q=2 magnetic surfaces. It is speculated that localized beams of the suprathermal electrons can be induced during magnetic reconnection around the X points of the m=1, n=1, and m=2,n=1 magnetic islands. Numerical simulations indicate that generation and losses of the suprathermal electrons in plasma with high density depend critically on the rate of the reconnection and amplitude of the magnetohydrodynamic modes.

Measurement and Control of the Growth of Rotating Magnetic Islands by Electron Cyclotron Current Drive in a Tokamak

The design of a feedback system based on the application of electron cyclotron (EC) current drive in high-density tokamak plasmas is presented. A realistic evaluation is given of the current drive produced by high-power gyrotron tubes for EC-wave beams injected from the low field side in the O-mode polarization. The driven current, calculated using a relativistic analytical formalism along the ray trajectories, is inserted in a consistent theory of nonlinear tearing-mode evolution. Results of modeling calculations are presented, showing the possibility of controlling by means of a digital system the evolution of the width of rotating magnetic perturbations by phase-locked modulation of the gyrotron power source.

A one-dimensional model for radiative thermal equilibrium and stability of the tokamak edge plasma

The multifaceted asymmetric radiation from the edge (MARFE) instability observed in the peripheral radiative layer in the high-density tokamaks is considered for a spatially smooth profile of radiative-loss function in equilibrium. The instability which is driven by the impurity radiation and opposed by the finite thermal conductivity, was considered previously by Drake [J. F. Drake, Phys. Fluids 30, 2429 (1987)] as a surface wave on a sharp radiative layer (step-function-like) at the edge. A novel exponential model is presented for the radiative loss function and it is shown that the two crucial features of MARFE: (i) stability to poloidally symmetric perturbations and (ii) a threshold value of radiative loss for the growth of asymmetric perturbations, explained earlier by Drake’s theory, are contained in this model of radiative equilibrium, also. Since, from the earlier theory, these features appear to be equilibrium specific and related to the form of radiative-loss function, the present work broadens the class of radiation profiles which can explain the MARFE, by recovering all the old results by an entirely different mathematical treatment.

Δn = 0 N-shell emission of rare earth ions (Z = 59 to 70) emitted from low and high density tokamak and laser produced plasmas

A discussion of the line and unresolved transition arrays emission of rare earths ions isoelectronic with PdI to CuI emitted from low (∼ 1013 cm−3) and high (∼ 1020 cm−3) density tokamak and laser produced plasmas is presented. The spectra of prasaeodymium, europium, gadolinium, dysprosium and ytterbium have been excited in the TEXT tokamak plasma and recorded in the 50-200 Å range by means of a photometrically calibrated, time-resolving multispectral grazing incidence spectrograph. The prasaeodymium, samarium, gadolinium, dysprosium and ytterbium spectra obtained by focussing an Nd-glass laser beam on solid targets were recorded in the same spectral range on photographic plates. The relationship between the main plasma parameters, electron temperature and density, and the soft X-ray emission of rare earths is analyzed and discussed.

Recent trend of plasma diagnostics using scattering/interferometer technique.

Recent research and development on plasma diagnostics related with the scattering/ interferometer technique are surveyed. In particular, the alpha particle dianostics using the collective scattering with the X-mode radiation and the polarimeter for measuring the current density profile in high density and field tokamak plasmas are described.

Determination of Pitch Angles of Magnetic Field Lines in Tokamak with Obliquely Injected Helium Beam

The pitch angles of magnetic field lines or q-profile in a high-density tokamak can be determined by the new method proposed here. Blobs of monoenergetic helium beam with different velocities are injected into the tokamak obliquely to the toroidal direction on a poloidal plane χ. Produced He+ ions move nearly along magnetic field lines, and their front emitting a Hell line crosses a different sight plane and forms a curve on it. An optically measured shift of the curve from χ leads to a distribution of pitch angle after correction for the perpendicular drift motion of He+ ions, using the velocity dependence of the drift.

The O-X-B mode conversion scheme for ECRH of a high-density Tokamak plasma

A method to apply electron cyclotron resonance heating (ECRH) to a Tokamak plasma with central density higher than the critical density for cut-off of the ordinary mode (O-mode) has been investigated. This method involves two mode conversions, from an O-mode via an extraordinary mode (X-mode) into an electron Bernstein mode (B-mode). Radial profiles for the power deposition and the wave-drive current due to the B-waves are calculated for realistic antenna radiation patterns with parameters corresponding to the Danish DANTE Tokamak and to Princeton's PLT.