The analytical solution in toroidal coordinates of the Grad Shafranov equation has been at the origin of the tokamak breakthrough in the fusion development. Unfortunately, the standard toroidal ...coordinates have a circular poloidal section, which does not fit the elongated cross-section of the present tokamak experiments. In axisymmetry, the vacuum Grad Shafranov equation coincides with the Laplace equation for the toroidal component of the vector potential. In the present paper the solutions for the Laplace equation and that for the vacuum Grad Shafranov equation are tackled in the elliptical prolate toroidal cap-cyclide coordinates framework. The following report of the geometrical properties and of the metric of these coordinates allows us to work out the analytical solution of both equations in terms of the Wangerin functions.
The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been ...proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated.
The DTT facility is designed to study, within the European Fusion Road Map, the completely integrated (physics-technology and bulk-edge) power exhaust problems. The scientific program is ...characterized by the tight balance between clear guidelines and the “flexibility” to tackle all the present scientific questions as well as the ideas possibly arising along the years of its operative life; in particular, it will be ready to answer to the necessities arising from both ITER operation and/or DEMO design. The paper discusses the guideline of DTT operations program to be carried out during its operative life, planned in about thirty years.
The Divertor Tokamak Test (DTT) facility has been launched to investigate alternative power exhaust solutions for DEMO. DTT should offer sufficient flexibility to be able to incorporate the best ...candidate divertor concept (e.g. conventional, Snowflake, Super-X, Double Null, liquid metals). In this paper, the revised up-down symmetric DTT device is presented. The up-down symmetrisation of the device makes it possible to have the reference values of the plasma current up to 5.5 MA and, at the same time, it has an impact on the costs, for which a slight revision of the main parameters has been considered. The DTT alternative magnetic configurations, such as Double Null, SnowFlake, Super-X, Double Super-X and Single Null with reverse triangularity, guarantee suitable constraints on the plasma-wall distance and the plasma elongation. The feasibility of the configurations is evaluated in terms of maximum vertical forces and currents on the PF coils along the scenarios.
•Design review of the Italian Divertor Tokamak Test facility: engineering aspects.•Rationale for design choices and main differences from the original proposal.•Main goal of DTT: explore and qualify ...alternative power exhaust solutions for DEMO.•Edge conditions similar to DEMO in terms of dimensionless parameters.•Test of alternative divertor solutions in integrated scenarios.
This paper presents the engineering aspects of the design review of the Italian DTT (Divertor Tokamak Test facility), illustrating the rationale for the design choices and focusing on the main differences with respect to the original proposal.
•Present for the first time the design of DTT divertor.•Presents a comparison between standard SN configuration and ADCs.•Presents possible divertor scenarios for DTT.•Presents expected power exhaust ...performances of the designed DTT divertor.•Presents the impact of the long legs configuration on power exhaust performances.
The standard positive triangularity H-mode tokamak operation with Single Null Divertor (SND) configuration and tungsten monoblocks targets can face some difficulties in providing a solution scalable towards the realization of the fusion reactor. To provide a safer solution, the adoption of alternative divertor magnetic configurations (ADCs) has been considered to provide on the targets a stationary heat load of less than 10 MW/m2 presently considered an upper technological limit for a nuclear fusion reactor and to reduce enough plasma temperature to avoid tungsten sputtering and its possible plasma contamination. Additionally, different plasma scenarios have been considered to avoid the huge transient energy released due to the type-I ELMs of the high confinement H-mode operation.
The new high field superconducting divertor tokamak test facility (DTT) 1 is presently under construction to specifically study power exhaust solutions in reactor relevant regimes. The first DTT divertor will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes. To test and compare a wide set of different power exhaust solutions, the divertor is being designed to be compatible with the standard single null (SND) configuration but also with some ADCs, like the X divertor (XD) and the hybrid Super-X/long leg SN. Additionally, the l-mode negative triangularity (NT) operation is considered important to explore as a solution to avoid ELMs.
Two different closed divertors with the same grazing angle (α = 2°) for the reference SND configuration have been considered: a so called “narrow” one with normalized divertor parameters similar to those presently foreseen for DEMO divertor and the “wide” one with a 50 % wider target aperture which allows testing long legs configurations as well as snowflake divertor configurations (SFD). For the “wide” divertor different dome sizes have been also considered.
The comparison between different shapes and their optimization have been done with the 2D edge fluid-kinetic code SOLEDGE2D-EIRENE. The code easily manages all magnetic configurations and can resolve the heat load on all components thanks the mesh filling the whole divertor and vessel volume.
In a representative sub-set of all the possible magnetic equilibrium configurations that can be realized in DTT, the edge modelling shows a better performance of the “wide” divertor compared to the “narrow” one. In fact, the “wide” divertor allows to have lower plasma temperature at the targets in pure deuterium at low PSOL power and also lower impurity concentrations is required to achieve plasma detachment by impurity seeding at the reference full power DTT scenario. The results seem to indicate that the improvement due to the longer legs possible with the “wide” divertor can overcome its lower apparent closure compared to the “narrow” divertor solution.
The I-DTT tokamak has been analyzed by means of the integrated COREDIV code simulations when either Li or Sn are used as liquid divertor target materials. It has been found that power to divertor can ...be strongly mitigated with LMD. The reason is that the solution is determined by the LM divertor properties, leading to the requirements that the heat load to the liquid target is reduced below a threshold value. The threshold is due to the limits to the plasma contamination by the evaporated material. In the case of Li target, the limit is set to ∼8 MW/m2 and is achieved by strong Li radiation in the divertor (vapor shielding). For Li, there is a low density limit and solution is only achievable if the plasma density is high enough. The low density operation might be recovered if Kr seeding is applied.
For the tin liquid divertor, H-mode operation is possible with efficient reduction of the heat flux to the divertor (∼11 MW/m2) in the evaporation efficiency reduced mode of operation and with the separatrix density high enough. The heat load reduction can be even more efficient (∼2.5 MW/m2) in the regime with strong evaporation but in this case the H-mode operation might be a problem. It appears that Ne seeding can hardly solve the H-mode operation problem but Li seeding seems to be better solution. The operation with higher edge plasma densities alleviates difficulties with the H-mode operation of liquid tin divertor.
Neutronics study for DTT tokamak building Colangeli, A.; Villari, R.; Luis, R. ...
Fusion engineering and design,
September 2019, 2019-09-00, 20190901, Letnik:
146
Journal Article
Recenzirano
•Divertor Tokamak Test (DTT) has been presented as a facility aimed at testing various alternative divertor concepts.•The spatial distribution of the fluxes, spectra and doses inside and outside the ...Tokamak hall has been evaluated with MCNP 360° simplified model.•All analyses have been done using MCNP-5 Monte Carlo code and ADVANTG hybrid has been used to implement specific variance reduction techniques.•To assess the impact of the wall thickness on fluxes and dose rates outside the wall, five concrete wall thicknesses were simulated (150 – 250 cm).•The results of this study provides the preliminary assessment of the radiation level for licensing procedure required by the Italian regulation.
The Divertor Tokamak Test (DTT) facility is a project proposed by an Italian consortium aimed to test the physics and technology of various alternative divertor concepts in order to design a heat and power exhaust system able to withstand the large loads expected in the divertor of a DEMO fusion power plant. The DTT machine is expected to produce up to 1.0 × 1017 2.5-MeV neutrons per second through deuterium-deuterium (DD) reactions with a significant 14 MeV neutron production (1% of the total yield) due to triton burn-up during the high-performance phase. This work is devoted to a preliminary three-dimensional shielding study to optimize the DTT building for licensing purposes.
Neutron and gamma transport simulations were carried-out with MCNP5 Monte Carlo code using weight windows generated with ADVANTG hybrid code. Three-dimensional model of the DTT machine and building has been developed to assess the spatial distributions of the neutron and gamma fluxes, spectra and effective doses inside and outside the tokamak hall. The wall thickness (in the range 150–250 cm) has no impact on satisfying dose limits for the public, but is important when satisfying limits for workers.
The possible scenarios at full power of the DTT (Divertor Test Tokamak) device with the standard single null (SN) and X divertor (XD) configurations have been analysed for the aspect of safely ...handling the power to be exhausted on the divertor targets. In this conceptual design phase the computational tools have been chosen mainly on the basis of their simplicity and rapidity. The code COREDIV was used for a preliminary self-consistent description of the coupled edge-core system. Subsequently, a more punctual analysis has been carried out on the SOL with the TECXY code. COREDIV results show that operations without impurity seeding may be problematic in all scenarios, and especially at the higher densities where tungsten is virtually absent in the core and the core radiation very low. The main outcome of this study is that in term of global parameters little difference exists between the two configurations for low working densities. The reason is identified in the fact that the topology modifications occur in region where the dissipative processes, radiation inelastic collisions etc., are rather negligible to be enhanced at significant level. The situation shows different at higher density where the XD seems indeed to favour detached operations and strongly radiating regimes. This trend is reinforced by lowering the power entering the SOL and by faster cross-field diffusion. These very important regimes seem to be reachable by the advanced configurations of DTT, for some appropriate choice of the working parameters.
•A preliminary design of liquid Li pool-type divertor for DTT is presented.•The design includes an evaporation chamber (EC) and a differential chamber (DC).•The Li loop is analyzed with a ...self-consistent thermodynamic model.•The Li vapor shield is shown to be effective for plasma heat flux redistribution.•Low Li vapor flux to the main plasma is achieved thanks to the DC pumping action.
Considering that solutions for the steady-state power exhaust problem in future fusion reactors (e.g. DEMO) are not provided by present experiments and it is uncertain if they will be provided by ITER, because the expected heat fluxes, as well as the level of neutron irradiation, will be much higher, dedicated work packages are being devoted to this problem within EUROfusion and even a dedicated facility (the Divertor Tokamak Test − DTT) is being proposed in Italy. Among the possible options, a liquid metal (LM) divertor is being considered. The present work aims at developing a simple model of the LM loop in the case of a pool-type divertor, including the most important physical phenomena and allowing to roughly determine the operating range of the system, in terms of surface temperatures and vapor pressures. This work therefore sets a preliminary basis for the conceptual design of a LM divertor for the DTT facility. The model includes the incoming plasma heat load and a basic treatment of the interactions of the Li vapor with the plasma. The reduction of the Li vapor efflux due to ionization by the plasma is also taken into account. The model includes two chambers: a first divertor box, the evaporation chamber (EC), is open towards a second divertor box, the differential chamber (DC), which is in turn connected to the main plasma chamber (MC). The model is used to study the effectiveness of the LM vapor in radiating isotropically the parallel heat flux incoming in the divertor. The results indicate that the presence of the DC allows a significant reduction of the Li vapor efflux towards the MC, which in turn would imply a lower contamination of the core plasma.
