Spherical tokamaks (STs) exhibit significant promise as the foundation for compact fusion power plants, offering reduced aspect ratios and enhanced plasma performance that can potentially lower ...capital costs compared to conventional tokamak designs. The key to achieving an optimal design lies in understanding the sensitivity of the fusion power plant to plasma energy confinement times. However, due to the intricate nature of transport physics and the scarcity of data on highly radiative plasmas required for power plants, extrapolating performance from existing machines introduces substantial uncertainties. In this study, we employed the world-leading fusion power plant systems code, PROCESS, to explore the effects of different energy confinement time scalings on scoping and determining the design of a 1-<inline-formula> <tex-math notation="LaTeX">\text{GW}_{e}</tex-math> </inline-formula> net electric ST power plant. By comparing various commonly used scalings, we highlight the design impact of employing ST scalings versus those typically applied to conventional aspect ratios, considering both size and performance aspects. Our findings demonstrate that when allowed to freely optimize the choice of confinement scaling has negligible impact on the optimally found design point and is instead driven highly by engineering constraints. In a highly constrained scenario, the conventional IPB98(y,2) scaling consistently shows conservative values across a range of ST plasma performance scenarios. We recommend its utilization for future large design space exploration studies as a low-risk choice due to its intermediary performance between the broad scope of ST scalings and also as a proxy for addressing complex transport considerations in configuring initial ST concept designs.
The article begins with a brief review of the achievements of Russian tokamaks in the active period of their development from 1962 to 1973, under the leadership of academician L.A. Artsimovich. ...During these years, the following basic issues were solved: the equilibrium problem, the MHD stability of the plasma column, and creation of the hot plasma with intense DD neutron radiation. It was shown that the ion confinement in tokamaks is close to the neoclassical model, and the electron confinement is abnormal. It improves with the increasing frequency of collisions, the opposite of the case with ions, in what is known as the alternative model of confinement along the magnetic field. Finally, the first scaling law for the energy lifetime of plasma was obtained, which accurately predicted the plasma parameters of the next generation of tokamaks (the so-called T-4 scaling). The subsequent movement in this direction (the 'Artsimovich vector') led to the creation of DT reactors with a fusion power of up to 10 MW (TFTR, JET) and to the ITER project. The main objective of the further development of tokamaks is their transition to steady-state fusion operation, which is a prerequisite for their use in industrial power generation. This makes it necessary to re-evaluate the achievements and obstacles that have to be overcome. The first limitation which thus arises is the so-called PH/S limit, which limits the value of the plasma heating power in a tokamak, as well as the discharge duration t (the 'TRIAM vector') in current tokamaks ( t ~ 1/(PH/S)1.7). Analysis of the existing experimental data shows that the most probable reason for the limitation of PH in existing tokamaks is the breakdown of the plasma sheath in the places of direct contact of the plasma with the wall. The reason for limiting the discharge duration t may be the gradual accumulation of the erosion products in the contact zones of the plasma with the tokamak first wall, which can facilitate such a breakdown. Creating a closed circulating lithium flow between the first wall and plasma is proposed as the solution to the problem of accumulation of the products of the first wall erosion. Preliminary studies (appendices A and B) have shown that the undesirable accumulation of tritium in the protective lithium films can be avoided if the temperature of the wall of the tokamak discharge chamber does not exceed 400 °C.
30 Jahre ASDEX Upgrade Kallenbach, Arne
Physik in unserer Zeit,
05/2021, Letnik:
52, Številka:
3
Journal Article
ZusammenfassungDer Garchinger Tokamak ASDEX Upgrade ging vor dreißig Jahren in Betrieb. Seither wurde die Anlage, insbesondere ihre Gefäßeinbauten, Plasma‐Heizsysteme und Diagnostiken, beständig ...erweitert und verbessert. Mit dem Wechsel von einer mit Graphit zu einer mit Wolfram bedeckten Gefäßwand mussten viele Plasmaszenarien überarbeitet werden. Hierbei wurde auch die Kühlung des Divertorplasmas durch zusätzlich eingebrachte Strahlungsverluste entwickelt, die auf ein künftiges Tokamak‐Fusionskraftwerk übertragen werden kann. In vielen Belangen kann ASDEX Upgrade als „Blaupause“ eines solchen Tokamak‐Fusionskraftwerks angesehen werden. Eine große Anzahl von Musterentladungen erlauben zusammen mit neu entwickelten Computercodes verlässliche Vorhersagen für ein Kraftwerk.
A 1-5/8" liquid stub tuner (LST) with effective impedance matching for ion cyclotron RF signal transmission is presented. The main objective is to maximize the power transmission from RF source to ...antenna for Steady-State Superconducting and Aditya Tokamaks. The design and characterization of an LST for varying susceptance from 10 to 100 MHz are described here. The difference in the dielectric constants of air and liquid makes the wavelengths different in these two mediums. This difference in wavelengths of RF waves is utilized for changing the impedance of LST. The LST's experimental susceptance is confirmed by CST simulation and MATLAB analysis. The single-stub matching mechanism is implemented using the validated approach. The theoretical research indicates the LST's mechanical benefits over traditional ones. Instead of finger contact-based sliding mechanics, the LST uses liquid height movement. It avoids sparks from high-current finger connections in traditional air dielectric stub tuners' moving plungers. This idea for impedance matching overcomes the constraints of traditional stub tuners where spark gets developed between moving plunger and coaxial conductors.
Due to restricted space for a full-sized solenoid, compact tokamaks and spherical tokamaks such as the ST40 and the next-step ST80 device planned by Tokamak Energy, Oxfordshire, U.K., could ...considerably benefit from solenoid-less plasma startup capability. Transient coaxial helicity injection (CHI), a method first developed on the small helicity injected torus-II (HIT-II) experiment and then validated on national spherical torus experiment (NSTX), is a method to initiate an inductive-like tokamak plasma discharge without reliance on the central solenoid. In both these devices, toroidal ceramic insulators were used to electrically separate the inner and outer vessel components so that the magnetic flux that initially connects the inner and outer vessel components could be grown into the vessel using J <inline-formula> <tex-math notation="LaTeX">\times</tex-math> </inline-formula> B forces to generate a closed magnetic field line configuration. In reactors, the installation of large toroidal insulators as part of the vacuum vessel boundary may not be feasible. To address this design requirement, a first of its kind, floating double-biased reactor-relevant CHI configuration is being developed for Pegasus-III. A particularly challenging requirement for a successful transient CHI discharge generation is the need to reduce the amount of fuel gas injected to the levels normally used for discharge initiation using the inductive solenoid. While higher levels of gas injection ease the plasma breakdown requirements, it also favors discharge initiation outside the CHI injector region. This stringent requirement generally requires that a small gas plenum be located close to an appropriately designed gas injection manifold in the injector region in a high gas conductance configuration. Plasma startup using transient CHI and the design aspects for a transient CHI gas injection system for the Pegasus-III geometry are described.
In metallic fusion devices, parasitic light originating from multiple reflections on the wall is a major problem for the interpretation of optical diagnostics. Strong stray light affects several ...optical diagnostics in ITER. One possibility to cope with this reflected light is to use photonic simulation, which can accurately predict the behavior of light within complex 3D geometry. A prerequisite is to get a good description of the reflection model, represented by the Bidirectional Reflectance Distribution Function (BRDF), based on optical measurements of in-vessel materials. To avoid complicated measurements using goniophotometer to get the BRDF, one possibility is to link surface optical properties and topography characteristics, such as roughness measurements, for example, using the classical Bennett’s formula. Measurements were performed using two experimental goniophotometers to fully characterize the BRDF of tungsten samples with different roughness values. Surface topography was measured using a three-dimensional laser scanning confocal microscope. Several parameters were extracted from these measurements including the arithmetic average roughness (Ra), the root mean square roughness (RMS), the Surface Inclination Angle Distribution and furthermore its mean value δm and the power spectral density (PSD). The correlations of BRDF model parameters deduced from the measurements are compared with the previous topographic parameters. The initial results on several tungsten samples show that Ra, which is the usual measure of surface roughness, is not the most suitable metric to link with the reflection behavior of the surface. In contrast, the PSD and the surface inclination angle are interesting metrics for describing the reflected light.
•Phong parameters ks and kd follow the specular and diffuse reflectance, respectively•Ra, RMS and Rs are not suitable metric to link surface reflection and topography•δm and σrel and RMSs are good topography metrics to relate to Phong model parameters
Abstract The two best performing pulses of the so called ITER-Baseline scenario ( I p = 3.5 MA and P in ≈ 35 MW) of JET-ITER like wall, one in deuterium (D) the other in deuterium–tritium (D–T) ...plasma are examined and compared in this study. Generally, the D–T Baseline pulses exhibit an electron density level higher than the D pulses and the plasma energy is higher than in the comparable D pulses by up to 20%, reaching about 12 MJ in the pulse studied here. In contrast with the D pulses, the D–T pulses are often characterised by the increase in time of the radiated power in the mantle region (0.70 < ρ < 0.95), which may lead to the loss of the edge localised mode activity when the threshold H–L transition power is approached and to the subsequent plasma disruption due to excessive radiation. In this study we try to identify the physical mechanisms responsible for this behaviour using the available experimental data (principally the total radiated power from the bolometry) and the results of the fluid COREDIV model (1D in the core, 2D in the scrape-off-layer (SOL)), self-consistent with respect to core-SOL and also to main plasma-impurities. In fact, the loss of power caused by impurity radiation affects the temperature profile and finally the power to the divertor plate. The electron density and temperature profiles are numerically reconstructed as well as the radiated power density profiles, indicating no major difference in impurity transport in D and D–T. Indeed, the impurity transport coefficients used in COREDIV to match the experimental radiated power profiles are similar in the two pulses. The computed tungsten sources and densities are lower in the D–T pulse and the divertor impurity retention capability is a little better in the D–T pulse, indicatinga stronger collisional drag force in the SOL. The higher electron density and the broadening of its profile are the main cause of the observed increase of the radiated power in the D–T pulse.
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