Divertor Tokamak Research Articles

Installation, thermal curing, qualification testing of divertor and position control coils in ADITYA-U tokamak

The former ADITYA, a medium-sized tokamak with a limiter configuration was upgraded to ADITYA-U tokamak with divertor configuration. Two pairs of new divertor coils, a single pair of auxiliary divertor coils and position control coils have been introduced in ADITYA-U tokamak to achieve shaped plasma operation using the existing Toroidal field, Ohmic transformer and vertical field coils. Currently, copper-based continuous transposed conductor (CTC) has been introduced for in-situ winding of the coils. Coil insulation materials are selected to withstand high current (15 kA), high voltage (5 kV) and sustain high temperature (120 °C) during the experiment. The primary challenge was to install the coil with the bus bar using the same conductor without any joints. The design of new divertor coils mainly includes electrical and thermal considerations within the limited space available for installation. A detailed description of the installation of coils, insulation fabrication, insulation curing process and testing of the divertor coils is presented in this paper.

Edge-localized mode mitigation enabled by active control of pedestal density gradient with new EAST tokamak divertor

Edge-localized mode mitigation enabled by active control of pedestal density gradient with new EAST tokamak divertor

Cross-code comparison of the edge codes SOLPS-ITER, SOLEDGE2D and UEDGE in modelling a high-power neon-seeded scenario in the DTT

Abstract In this work, we cross-compare the state-of-the-art edge plasma codes SOLPS-ITER, SOLEDGE2D, and UEDGE in a reactor-relevant neon-seeded Divertor Tokamak Test (DTT) scenario at nominal power, extending the simplified test-bed of Moscheni et al. (2022). Converged solutions targeting the same separatrix density and radiated power are obtained by adjusting the pumping albedo and the neon puffing rate. This higher-power scenario is generally characterised by substantial disagreement between the three codes, up to 78-178% in peak heat fluxes. Discrepancies found in Moscheni et al. (2022) are indeed exacerbated, and new ones arise. Underlying causes include the over-penetration of neutrals implied by the unified ion-neutral temperature of UEDGE (observed in Moscheni et al. (2022)), here resulting in a 38-114% over-estimation of core plasma densities. The particular set of EIRENE atomic-molecular reactions adopted is found to stiffly restrict the achievable code solutions, which results in the predicted effective charge Z eff changing from ∽5 to ∽8 at the outer mid-plane separatrix. The strong link between Z eff, main ion density and unified ion temperature emphasises the need of proper assessments of impurity cross-field transport, with its implications on core contamination and wall erosion. The advantages of extended plasma meshes are found to come with associated modelling intricacies, yet to be fully characterised but seemingly impacting on the activity around the secondary X-point of single-null magnetic topologies. An appreciable impurity particle imbalance, generated by the neon ion density floor, is noted—speculatively contributing to SOLEDGE2D's different radiation emission distribution, and expected to be even more deleterious for high-Z impurities in all the codes. Potential drivers of further discrepancies are the different Braginskii formulations of collision times and momentum sources in presence of impurities, and the SOLPS-ITER extra terms gaining importance around the detachment front. Outstanding questions unanswered in this work prompt further investigations.

Divertor Tokamak Test: Impact of NBI shine-through and beam-plasma interaction on Divertor Tokamak Test facility

IntroductionIn this work, we aim to explore numerically the behavior of beam energetic particles in the Divertor Tokamak Test (DTT), a superconductive device equipped with a Neutral Beam Injection (NBI) system capable of injecting neutrals up to 510 keV.MethodWe explore beam ionization and beam slowing down for different DTT plasma scenarios. Numerical simulations are performed using the ASCOT suite of codes, including a wide-range scan of plasma density and beam injection energy. For different plasma conditions, we estimate shine-through losses, including the heat fluxes on the first wall thanks to dedicated particle tracing simulations. Orbits of newly-born fast ions are characterized by means of the constant of motion phase space, showing how trapped energetic particles’ population and prompt losses change with plasma density and NBI energy.Results and discussionSlowing down simulations show that NBI injection at 510 keV is well coupled to DTT plasmas. DTT NBI will be one of the sources of auxiliary ion heating, with an absorbed power ratio of up to ∼50% depending on plasma and beam parameters. At low plasma densities, energetic particle confinement is less efficient, and NBI power and/or energy reduction is expected.

Lyman line opacities in tokamak divertor plasmas under high-recycling and detached conditions

Radiation transport simulations of the deuterium Lyman line series using EIRENE predict a factor of 2 increase of Balmer-α emission in the JET-ILW low confinement divertor plasmas in high-recycling and detached conditions caused by the re-absorption of Lyman-β emission. The photon tracing model in the EIRENE code is applied to SOLPS-ITER pure deuterium plasma solutions of JET-ILW to evaluate the contribution of Lyman-β capture towards the population density of D(p=3) and, consequently, the Balmer-α emission. The local population escape factors of Lyman-α and Lyman-β lines in the JET-ILW divertor show steep spatial gradients and indicate local regions with high opacity in high-recycling and detached conditions. The relative contribution of Lyman-β capture towards the population density of D(p=3) is significant in high-recycling and detached conditions. This observation is further corroborated by line-integrated calculations using Balmer-α diagnostic line-of-sights.

Correlation between non-ambipolar currents and divertor heat loads in the COMPASS tokamak

Abstract Electric current flowing onto the divertor of the COMPASS tokamak influences its heat loading. The current measured when a Langmuir probe is grounded to the divertor gives a local measurement of this heat loading, which, according to the classical theory of the Debye sheath, should be enhanced with respect to the case of locally ambipolar currents. The comparison of the calculated heat flux by probes with infrared thermography, when the influence of non-ambipolar currents is not considered, is grossly wrong; when the theoretical effect of non-ambipolar currents is included, however, the agreement is very good.

Detached Plasma Studies in GOL-NB with Extra Gas Injection

The magnetic system of an open trap usually includes expansion sections located between highfield magnetic mirrors and end surfaces that receive plasma. In the GOL-NB device, an arc plasma gun is located in one of the expanders, which creates a low-temperature starting plasma in the confinement area. The parameters of the surface plasma sheath affect the electrical connection of the confinement area with the walls and, thereby, affect the contribution of the line-tying effect to the plasma stability and the longitudinal energy losses from the trap. The experiments with additional hydrogen injection into the plasma gun were carried out at GOL-NB. We observed a radiating plasma formation detached from the surface, which visually corresponds to that in radiating divertors in tokamaks. In both standard and detached modes, decaying plasma existed near the receiving electrodes during the entire observation time after the discharge current was terminated. In the central trap of GOL-NB, some structures in the Fourier spectrogram of magnetic fluctuations manifest earlier in the detachment mode than in the standard mode and have lower frequencies. We associate these structures with the onset of interchange-like modes due to the loss of plasma stabilization by the line-tying to the conducting ends. The observed plasma response to the additional gas supply confirmed our understanding of the line-tying effect as the main factor stabilizing the plasma core in the initial phase of density accumulation in the central trap.

Divertor Tokamak Test facility project: status of design and implementation

Abstract An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, XD divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

A long-pulse small edge-localized-mode high-confinement plasma with detachment feedback control by floating potential in an experimental advanced superconducting tokamak in a metal wall environment

Abstract One of the key challenges facing the magnetic fusion research is to demonstrate the compatibility between high confinement and radiative divertor in long-pulse discharges with a metal wall environment. A small edge localized mode (ELM) high confinement plasma is obtained in the upgraded lower divertor of Experimental Advanced Superconducting Tokamak (EAST) with an energy confinement factor H98~1.1 and a Greenwald density fraction fGW ~ 0.65 maintained for 26 s, and periodical detachment is achieved through active control of neon impurity seeding in this long-pulse discharge. For the divertor region, partial detachment is achieved periodically on the outer divertor target plates with the plasma temperature near the outer strike point decreased to below 5 eV and peak surface temperature on the outer divertor target plates maintained below 350C. The peak heat flux of the lower outer divertor decreases significantly and its profile along the target becomes very flat in the detached state. Two low-frequency (<10 kHz) fluctuations that are related to rippling mode caused by a resistive instability appear in the detached state. For the pedestal region, the electron pressure profile is flatter and ELM amplitude is smaller in the detached state than that in the attached state. Edge coherent mode (ECM) appears in the attached state and disappears in the detached state. To achieve this experimentally, a new impurity seeding feedback control scheme is applied, where the floating potential measured by divertor Langmuir probes is used as the feedback sensor, which is more reliable in the long-pulse discharges with high heat fluxes, thus more suitable for application in future devices. This work provides a new approach for the actively controlled radiative divertor as a solution to the divertor heat loads of the future fusion reactors.

Physics basis for the divertor tokamak test facility

Abstract This paper is dealing with the physics basis used for the design of the Divertor Tokamak Test facility (DTT), under construction in Frascati (DTT 2019 DTT interim design report (2019)) Italy, and with the description of the main target plasma scenarios of the device. The main goal of the facility will be the study of the power exhaust, intended as a fully integrated core-edge problem, and eventually to propose an optimized divertor for the European DEMO plant. The approach used to design the facility is described and their main features are reported, by using simulations performed by state-of-the-art codes both for the bulk and edge studies. A detailed analysis of MHD, including also the possibility to study disruption events and Energetic Particles physics is also reported. Eventually, a description of the ongoing work to build-up a Research Plan written and shared by the full EUROfusion community is presented.

Design of an infrared/visible endoscope system for HL-3 divertor observation

This paper presents the design and test results of an infrared/visible endoscope system for diagnosing divertor in HL-3 tokamak. The system is designed to view the plasmas tangentially with multi-imaging systems including two infrared imaging systems and three visible imaging systems. Several filters based visible imaging measurements are thus available simultaneously for divertor physics investigation. The performance of the optical design is explored, and the manufacture of the system and the test results are presented. The characteristics of the advanced divertor configurations, such as multi-strike points and two X-points, are observed clearly with the endoscope system.

Armor Thickness Assessment for the Divertor Tokamak Test Facility (DTT) Divertor Targets

Armor Thickness Assessment for the Divertor Tokamak Test Facility (DTT) Divertor Targets

Status of Design of the First Wall Supports of the Divertor Tokamak Test Facility

Status of Design of the First Wall Supports of the Divertor Tokamak Test Facility

Negative triangularity scenarios: from TCV and AUG experiments to DTT predictions

Abstract Experiments, gyrokinetic simulations and transport predictions were performed to investigate if a negative triangularity (NT) L-mode option for the Divertor Tokamak Test (DTT) full-power scenario would perform similarly to the positive triangularity (PT) H-mode reference scenario, avoiding the harmful edge localized modes (ELMs). The simulations show that a beneficial effect of NT coming from the edge/scrape-off layer (SOL) region ρ tor > 0.9 is needed to allow the actual NT L-mode option to perform like the PT H-mode. Dedicated experiments at TCV and AUG, with DTT-like shapes, show an optimistic picture. In TCV, experiments indicate that even with the relatively small triangularity of the DTT NT scenario, a large beneficial effect of NT comes from the plasma edge and SOL, allowing NT L-modes to outperform PT L-modes with the same power input, reaching the same central pressures as PT H-modes with twice as much applied heating power. For AUG, NT plasmas go into H-mode more easily than for TCV, but always present much smaller pedestals compared with PT plasmas with the same input power, showing a much weaker or absent ELM activity. However, NT has a smaller beneficial effect for AUG than for TCV, with NT pulses outperforming PT pulses with the same input power only for an ECRH-only case with relatively low input power. For the considered AUG cases, PT pulses perform better than NT ones at higher ECRH power or with mixed NBI and ECRH power. Based on this analysis, the NT option is a viable alternative for the DTT full power scenario, providing high performance plasmas with reduced or absent ELMs.

Towards the DTT configuration management platform architecture

Effective project management methods, tools and working practices shall be applied to facilitate the communication and collaboration among the different institutions involved in the Divertor Tokamak Test (DTT) project. This paper deals with the definition of the configuration management platform architecture enabling technical integration of the DTT system. The first step consists in the identification of main requirements the platform should satisfy, considering the multidisciplinary domains and the geographically dispersed working teams characterizing the nuclear fusion sector and the maturity level of the specific project. Main characteristics of the most advanced Product Lifecycle Management (PLM) tools are identified, their limits and benefits are evaluated, and the suitable PLM platform is selected, satisfying the DTT project needs. The definition of the architecture of the configuration management platform for the DTT project aims at implementing the DTT assembly model in a unique environment able to exchange models and data even developed outside the platform ensuring the congruence of the design, the traceability of design changes and the adoption of a proper Systems Engineering approach.

Sputtering Coefficients of Beryllium and Tungsten by Various Atoms from Hydrogen to Tungsten

Using computer simulation, the sputtering coefficients of Be and W targets, promising materials for the first wall and divertor in the ITER tokamak, are calculated in a wide range of incident atom energies 10–100 000 eV. The following atoms were chosen as projectiles: H, D, T, He, Be, C, N, O, Ne, Ar, W. A strong influence of the surface profile on the results obtained is shown. The limiting cases of a planar potential barrier (smooth surface) and a spherical potential barrier (rough surface) are considered. Data on the average energy and angular distributions of sputtered atoms were obtained, which are necessary for calculating the influx of impurities into the tokamak plasma. The influx of wall material atoms into the ITER tokamak plasma is estimated when the wall is sputtered by flows of fast deuterium and tritium atoms leaving the plasma.

Experimental and simulation study of argon helicon discharge in multiple plasma simulation linear device (MPS-LD)

Abstract To obtain a high-parameter plasma in the target region of a multiple plasma simulation linear device and to realize the experimental simulation environment of tokamak divertor plasma, experimental and numerical simulations of argon helicon discharge are carried out. Langmuir probes are used to diagnose the electron density (ne ) in the source and target regions with different experimental parameters (magnetic field, radio frequency power, puffing flow rate). A three-dimensional discharge model is developed using drift-diffusion equations of electron density and electron energy with the aid of COMSOL. Helicon discharge with a long straight plasma beam and a bright blue core is experimentally achieved. The simulation and experimental results are compared, validating the model. The corresponding spatial ne distribution is obtained, and the dependence of ne on the main experimental parameters is confirmed. The energy conversion relationship between the helicon and plasma is found. Helicon waves prefer to transfer energy to the plasma in the source region, where ne is significantly increased. This results in a strong ne gradient, which acts as a barrier to prevent the propagation of helicon waves. Therefore, localized standing helicon waves are formed, which limits the increase in plasma density in the target region. By increasing the magnetic field strength (B < 1500 G) and RF power (P < 1500 W), ne in the source region can be increased, but they have little effect on ne in the target region.

Experiments on plasma detachment in a V-shaped slot divertor in the DIII-D tokamak

Experiments in DIII-D demonstrate that the upstream plasma density to detach an un-pumped slot divertor is similar for a V-shaped and a flat-end slot, despite significantly higher neutral pressure in the V-shaped slot and in contrast to SOLPS-ITER predictions. The detachment threshold can be reduced by using in-slot instead of main-chamber gas fuelling or by placing the strike point on the inner slanted slot baffle instead of the slot end, as described by simulations with full drift physics. When increasing the plasma line-averaged density (without extrinsic impurities), the transition to detachment in DIII-D slot divertor is sharp and requires a high value of plasma density with the ion B→×∇B drift into the slot, whereas it is smooth and requires a lower value of plasma density with the opposite drift direction, in accord with detachment experiments in the DIII-D open lower divertor. Unique experiments on DIII-D and comparison to advanced simulations expand the scientific understanding of slot-shaped divertors, considered highly desirable for next step fusion devices.

Conceptual design of a Doppler spectrometer for 102m/s cross-field flows in tokamak divertors.

The cross-field transport in the scrape-off-layers (SOLs) and divertors in tokamaks is of a similar size to the poloidal component of the parallel flow, thereby significantly impacting the plasma transport there. However, its direct observation has been challenging because the drift velocity (102-103 m/s) is significantly below the detection limit of conventional diagnostics. To realize cross-field ion flow measurement, a variety of systematic uncertainties in the system must be narrowed down. Here, we develop a conceptual design of the Doppler spectrometry that enables us to measure the impurity flows with 102-m/s accuracy based on an in situ wavelength-calibration technique developed in the astrophysics field, the iodine-cell method. We discuss its properties and applicability. In particular, the scaling relation between wavelength accuracy and various spectroscopic parameters is newly presented, which suggests the high importance of the wavelength resolution of the system. Based on transport simulations for the JT-60SA divertor, the feasibility of the system is assessed.

Calculation of toroidal Alfvén eigenmode mode structure in general axisymmetric toroidal geometry

A workflow is developed based on the ideal MHD model to investigate the linear physics of various Alfvén eigenmodes in general axisymmetric toroidal geometry by solving the coupled shear Alfvén wave (SAW) and ion sound wave (ISW) equations in ballooning space. The model equations are solved by the FALCON code in the singular layer, and the corresponding solutions are then taken as the boundary conditions for calculating parallel mode structures in the whole ballooning space. As an application of the code, the frequencies and mode structures of toroidal Alfvén eigenmode (TAE) are calculated in the reference equilibria of the Divertor Tokamak Test facility with positive and negative triangularities, respectively. As typical result for reactor relevant plasma conditions, which are strongly triangular in the outer core region where magnetic shear is of order unity, we show that the triangularity effect on TAE is generally small. Furthermore, by properly handling the boundary conditions, we demonstrate finite TAE damping due to coupling with the local acoustic continuum and find that the damping rate is small for typical plasma parameters.