Aiming at achieving steady state operation of Tokamak, the divertor is supposed to exhaust particle and heat flux. Plasma configuration is one of the most important drivers for divertor design determining the position of stricken point and heat flux intensity. In order to provide flexibility with plasma configurations, a continuous V shape divertor structure with consistent heat removal...
The Divertor Tokamak Test facility (DTT) [1] is a fusion device currently under construction in ENEA Frascati. Its main scientific goal is to investigate advanced solutions for the heat exhaust in future fusion power plants, such as DEMO. A key step in the success of DTT is the development of integrated plasma scenarios where good core performance is achieved together with acceptable...
Experimental observations on TCV [1] and DIII-D [2] have shown that negative triangularity L-Mode discharges can exhibit H-mode grade confinement, opening the possibility for high confinement reactors that side-step the challenges associated with H-mode such as ELMs, narrow scrape-off layer widths, and density control. To ensure safe power exhaust that protects the plasma facing components,...
Injection of boron (B) and boron nitride (BN) in powder form into the upper closed divertor at DIII-D showed a substantial drop in divertor electron temperature from 30 eV to below 5 eV, increase in divertor neutral compression by up to an order of magnitude, and transition into stable detachment [1]. A decrease in wall fueling, main chamber neutral pressure, and the reduction of oxygen,...
Divertor leg length requirements for testing divertor detachment compatibility with a hot X-point plasma and robust H-mode pedestal is examined with DIII-D data and modeling. Poloidal Te gradients consistent with this requirement are found to be determined by convective energy transport and particle balance constraints. These considerations provide estimates of divertor leg length requirements...
Future large, high power fusion devices such as ITER and its successors will require strong dissipation of the plasma energy, momentum and particles reaching open field lines before reaching the surfaces of the divertor; detachment will play a dominant role in that dissipation. This contribution presents initial observations of divertor detachment from the MAST Upgrade spherical tokamak in...
Divertor recycling control is enabled by a new tungsten lower divertor on EAST, featured by a right-angled closed corner joining the vertical and horizontal target plates. ELM mitigation is observed as the outer strike point moves from the vertical target to the horizontal target with a significant reduction of the pedestal density gradient. Furthermore, the new closed corner divertor exhibits...
Divertor detachment with medium-Z impurities seeded through gas puffing can entail radiating regions within the last closed flux surface. The lithium vapor box divertor seeks to detach via near-target low-Z lithium evaporation with the result that such a radiating region does not form. We show SOLPS-ITER predictions for the effect of a lithium vapor box divertor on NSTX-U. Past work has shown...
Results from calculations of the neutron flux in the divertor cassette body for the Helium Cooled Pebble Bed and Water-Cooled Lithium Lead concepts are presented in this paper. For both cases under investigation, the same divertor model setup and designated DEMO neutron source were utilized. Neutron transport calculations were performed using the MCNP6 and the FENDL-3.1 nuclear data library....
Within the roadmap of the EU-DEMO reactor design, the divertor design is being developed in the framework of the EUROfusion Consortium Work package “Divertor”, sub-project DEMO (WPDIV-DEMO).
The ultimate objective of this sub-project is to deliver at least one holistic design concept and feasible technology options for the DEMO divertor and limiter eligible for the subsequent Engineering...
Time-dependent SOLPS-ITER simulations have been extensively applied to various problems such as actuator design, feedback control through system identification, and physical interpretation of dynamic problem. Most SOLPS-ITER simulations focus on steady-state solutions, and the EIRENE module, a neutral solver, couples in a time-independent mode assuming quasi-relaxation within time step of 1e-3...
Over the first phase of WEST exploitation, the full tungsten environment consisted in a mixture of coated tiles and actively cooled ITER-like monoblocks. Using a combination of LHCD and ICRH heating, diverted L-mode scenarios were extended up to 50 s and stationary heat loads reaching 6 MWm-2 were deposited on ITER-like monoblocks. Despite input power levels generally above expected thresholds...
