| 1. |
- Agredano Torres, Manuel, et al.
(författare)
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Coils and power supplies design for the SMART tokamak
- 2021
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 168, s. 112683-112683
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Tidskriftsartikel (refereegranskat)abstract
- A new spherical tokamak, the SMall Aspect Ratio Tokamak (SMART), is currently being designed at the University of Seville. The goal of the machine is to achieve a toroidal field of 1 T, a plasma current of 500 kA and a pulse length of 500 ms for a plasma with a major radius of 0.4 m and minor radius of 0.25 m. This contribution presents the design of the coils and power supplies of the machine. The design foresees a central solenoid, 12 toroidal field coils and 8 poloidal field coils. Taking the current waveforms for these set of coils as starting point, each of them has been designed to withstand the Joule heating during the tokamak operation time. An analytical thermal model is employed to obtain the cross sections of each coil and, finally, their dimensions and parameters. The design of flexible and modular power supplies, based on IGBTs and supercapacitors, is presented. The topologies and control strategy of the power supplies are explained, together with a model in MATLAB Simulink to simulate the power supplies performance, proving their feasibility before the construction of the system.
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| 2. |
- Biel, W., et al.
(författare)
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Development of a concept and basis for the DEMO diagnostic and control system
- 2022
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Ingår i: Fusion engineering and design. - : Elsevier. - 0920-3796 .- 1873-7196. ; 179
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Tidskriftsartikel (refereegranskat)abstract
- An initial concept for the plasma diagnostic and control (D&C) system has been developed as part of European studies towards the development of a demonstration tokamak fusion reactor (DEMO). The main objective is to develop a feasible, integrated concept design of the DEMO D&C system that can provide reliable plasma control and high performance (electricity output) over extended periods of operation. While the fusion power is maximized when operating near to the operational limits of the tokamak, the reliability of operation typically improves when choosing parameters significantly distant from these limits. In addition to these conflicting requirements, the D&C development has to cope with strong adverse effects acting on all in vessel components on DEMO (harsh neutron environment, particle fluxes, temperatures, electromagnetic forces, etc.). Moreover, space allocation and plasma access are constrained by the needs for first wall integrity and optimization of tritium breeding. Taking into account these boundary conditions, the main DEMO plasma control issues have been formulated, and a list of diagnostic systems and channels needed for plasma control has been developed, which were selected for their robustness and the required coverage of control issues. For a validation and refinement of this concept, simulation tools are being refined and applied for equilibrium, kinetic and mode control studies.
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| 3. |
- Cecconello, Marco, et al.
(författare)
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Conceptual design of a collimated neutron flux monitor and spectrometer for DTT
- 2021
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Ingår i: Fusion engineering and design. - : Elsevier. - 0920-3796 .- 1873-7196. ; 167
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Tidskriftsartikel (refereegranskat)abstract
- A conceptual design and performance studies for a collimated neutron flux monitor and neutron spectrometer for the Divertor Tokamak Test (DTT) facility are presented. This study is based on the single-null divertor configuration and for “Half Power” and “Full power” scenarios with 15 MW of negative-ion NBI, 29 MW of ECH and 3 MW of ICRF heating with a maximum neutron yield of 1.5 × 1017 s−1. Fast ion distributions (both from auxiliary heating systems and fusion born) have been simulated in TRANSP/NUBEAM and the corresponding neutron energy spectra have been calculated using DRESS. Synthetic diagnostics have been implemented to determine the neutron fluxes and spectra at the detector location. Neutron emissivity profiles, plasma position, core ion temperature and the ratio of thermal and non-thermal D ion populations can be obtained with good accuracy and time resolution.
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| 4. |
- De Angeli, M., et al.
(författare)
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Cross machine investigation of magnetic tokamak dust : Morphological and elemental analysis
- 2021
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 166
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Tidskriftsartikel (refereegranskat)abstract
- The presence of magnetic dust can be an important issue for future fusion reactors where plasma breakdown is critical. Magnetic dust has been collected from contemporary fusion devices (FTU, Alcator C-Mod, COMPASS and DIII-D) that feature different plasma facing components. The results of morphological and elemental analysis are presented. Magnetic dust is based on steel or nickel alloys and its magnetism is generated by intense plasma material interactions. In spite of the strong similarities in terms of morphology and composition, X-ray diffraction analysis revealed differences in the structural evolution that leads to non-trivial magnetic responses.
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| 5. |
- Dittrich, Laura, et al.
(författare)
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Retention of noble and rare isotope gases in plasma-facing components-Experience from the JET tokamak with the ITER-like wall
- 2023
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 192
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Tidskriftsartikel (refereegranskat)abstract
- Plasma edge cooling, ion cyclotron wall conditioning and disruption mitigation techniques involve massive gas injection (by puffs or pellets) to the torus. A certain fraction remains in plasma-facing components (PFC) due to co-deposition and implantation. An uncontrolled release/desorption of such retained species affects the stability of plasma operation. The aim of this work was to determine the lateral and depth distribution of noble (3He, 4He, Ne, Ar), seeded (N2, Ne, Ar) and tracer gases (15N, 18O) in PFC retrieved from the JET tokamak with the ITER-Like Wall (JET-ILW) after three experimental campaigns (ILW-1, ILW-2, ILW-3). Results regarding the retention of those gases are shown as well as a comparison to the deuterium retention in the studied areas. Heavy ion elastic recoil detection analysis was used, being the only technique capable of detection and quantitative assessment of all elements, especially low-Z isotopes. Helium was found on the divertor Tile 5, locally up to 44.1015 3He cm-2 and 12.1015 4He cm-2, and on the limiters as well. Neon was found in two positions on the limiters, with up to 10.1015 Ne cm-2 and the 15N tracer on Be limiters exposed to ILW-3. A correlation of N retention with the N seeding rates for each campaign has also been found.
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| 6. |
- Doyle, S.J., et al.
(författare)
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Magnetic equilibrium design for the SMART tokamak
- 2021
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 171, s. 112706-112706
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Tidskriftsartikel (refereegranskat)abstract
- The SMall Aspect Ratio Tokamak (SMART) device is a new compact (plasma major radius Rgeo≥0.40 m, minor radius a≥0.20 m, aspect ratio A≥1.7) spherical tokamak, currently in development at the University of Seville. The SMART device has been designed to achieve a magnetic field at the plasma center of up to Bϕ=1.0 T with plasma currents up to Ip=500 kA and a pulse length up to τft=500 ms. A wide range of plasma shaping configurations are envisaged, including triangularities between −0.50≤δ≤0.50 and elongations of κ≤2.25. Control of plasma shaping is achieved through four axially variable poloidal field coils (PF), and four fixed divertor (Div) coils, nominally allowing operation in lower-single null, upper-single null and double-null configurations. This work examines phase 2 of the SMART device, presenting a baseline reference equilibrium and two highly-shaped triangular equilibria. The relevant PF and Div coil current waveforms are also presented. Equilibria are obtained via an axisymmetric Grad-Shafranov force balance solver (Fiesta), in combination with a circuit equation rigid current displacement model (RZIp) to obtain time-resolved vessel and plasma currents.
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| 7. |
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| 8. |
- Kovtun, Yu.V., et al.
(författare)
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ICRF plasma production in gas mixtures in the Uragan-2M stellarator
- 2023
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 194
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Tidskriftsartikel (refereegranskat)abstract
- This paper summarizes previous results and presents new studies on the ICRF plasma creation both in pure gases and gas mixtures. In all the experiments, the two-strap antenna was operated in monopole phasing with applied RF power of ∼100 kW. The research for plasma creation was carried out at RF frequencies near the fundamental hydrogen cyclotron harmonic.
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| 9. |
- Lee, S. E., et al.
(författare)
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Tritium distribution analysis of Be limiter tiles from JET-ITER like wall campaigns using imaging plate technique and β-ray induced X-ray spectrometry
- 2020
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 160
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Tidskriftsartikel (refereegranskat)abstract
- Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.
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| 10. |
- Mancini, A., et al.
(författare)
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Mechanical and electromagnetic design of the vacuum vessel of the SMART tokamak
- 2021
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 171
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Tidskriftsartikel (refereegranskat)abstract
- The SMall Aspect Ratio Tokamak (SMART) is a new spherical device that is currently being designed at the University of Seville. SMART is a compact machine with a plasma major radius () greater than 0.4 m, plasma minor radius () greater than 0.2 m, an aspect ratio () over than 1.7 and an elongation () of more than 2. It will be equipped with 4 poloidal field coils, 4 divertor field coils, 12 toroidal field coils and a central solenoid. The heating system comprises of a Neutral Beam Injector (NBI) of 600 kW and an Electron Cyclotron Resonance Heating (ECRH) of 6 kW for pre-ionization. SMART has been designed for a plasma current () of 500 kA, a toroidal magnetic field () of 1 T and a pulse length of 500 ms preserving the compactness of the machine. The free boundary equilibrium solver code FIESTA [1] coupled to the linear time independent, rigid plasma model RZIP [2] has been used to calculate the target equilibria taking into account the physics goals, the required plasma parameters, vacuum vessel structures and power supply requirements. We present here the final design of the SMART vacuum vessel together with the Finite Element Model (FEM) analysis carried out to ensure that the tokamak vessel provides high quality vacuum and plasma performance withstanding the electromagnetic loads caused by the interaction between the eddy currents induced in the vessel itself and the surrounding magnetic fields. A parametric model has been set up for the topological optimization of the vessel where the thickness of the wall has been locally adapted to the expected forces. An overview of the new machine is presented here.
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