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Self-replenishing liquid metal coatings are considered as a perspective divertor design able to withstand challenging particle and power loads of a fusion tokamak-reactor. Numerical modeling of the scrape-of-layer (SOL) plasma with advanced 2D codes, such as SOLPS, is necessary for developing of the ‘liquid-metal’ divertor. In this work we report on upgraded version of SOLPS 4.3 code liquid metal erosion module implemented earlier in our group and present results of simulations of T-15MD tokamak with Li-covered divertor plates. The erosion model includes all main processes Li erosion, i.e. physical sputtering, thermal sputtering, evaporation, and prompt redeposition. Unlike some other available implementations, Li atoms are considered in kinetic approximation in our version. A detailed analysis of Li erosion and flow in T-15MD configuration for various powers (6–12 MW) and H content in the SOL is presented. It is shown that the most of eroded Li particles are redeposited on the divertor targets, however, in some regimes absolute Li flow from the divertor is still large and might lead to significant main plasma dilution with Li. Vapor shielding effect is pronounced on both divertor targets in the most reasonable regimes providing low peak heat flux values at the target plates, less than 10 MW m ^−2 . The target erosion rate and surface temperatures are within limits of the most target designs. Moreover, in strongly shielded cases the target temperature can be even lower than the Li melting temperature meaning that external heating is required to keep Li flowing. Sensitivity analysis shows that our results are most sensitive to the target heat conduction parameters, i.e. the target thickness, outer surface temperature. It means that controlling the target cooling rate can be a useful tool for controlling the liquid Li divertor regime. Variation of the Li erosion rate parameters has little effect on the divertor performance.

Commercial fusion reactors will be faced with extremely high divertor target heat fluxes that will require mitigation. Simulations of detachment in an NSTX-U scenario projected to have 92 MW m ^−2 unmitigated peak target heat flux are presented, which reaches sub-10 MW m ^−2 target heat flux using a highly dissipating lithium vapor box divertor design. The lithium vapor box is a detached divertor design which employs lithium vapor evaporation and condensation to contain lithium below the X-point. Previous SOLPS modeling has indicated a lithium vapor box can reduce the heat flux down to 10 MW m ^−2 via simultaneous evaporation from the Private Flux Region (PFR) and the Common Flux Region (CFR) sides of the vapor box. It is found here that PFR evaporation has improved access to the separatrix leading to significantly more efficient power dissipation than CFR evaporation. Simulations of target evaporation with an evaporation distribution that is self-consistent with the temperature of a Capillary Porous System with Fast flowing liquid lithium could reach $n_\textrm{Li}$ / $n_\textrm{e} \sim$ 0.025–0.030 at the Last Closed Flux Surface (LCFS) depending on the liquid metal flow speeds and lithium sputtering yield, while PFR-side evaporation can reach acceptable heat fluxes with $n_\textrm{Li}$ / $n_\textrm{e} \sim$ 0.038 at the LCFS. However, PFR evaporator performance can be improved if the target is allowed to be hot enough such that it reflects lithium, reaching $n_\textrm{Li}$ / $n_\textrm{e} \sim$ 0.028 and reducing required lithium evaporation. Ultimately PFR evaporation and target evaporation are found to have similar ability to produce acceptable heat flux solutions with minimal upstream concentration.

One of the major design limitations for tokamak fusion reactors is the heat load that can be sustained by the materials at the divertor target. Developing a full understanding of how machine or operation parameters affect the conditions at the divertor requires an enormous number of simulations. A promising approach to circumvent this is to use machine learning models trained on simulation data as surrogate models. Once trained such surrogate models can make fast predictions for any scenario in the design parameter space. In future such simulation based surrogate models could be used in system codes for rapid design studies of future fusion power plants. This work presents the first steps towards the development of such surrogate models for plasma exhaust and the datasets required for their training. Machine learning models like neural networks usually require several thousand data points for training, but the exact amount of data required varies from case to case. Due to the long runtimes of simulations we aim at finding the minimal amount of training data required. A preliminary dataset based on SOLPS-ITER simulations with varying tokamak design parameters, including the major radius, magnetic field strength and neutral density is constructed. To be able to generate more training data within reasonable computation time the simulations in the dataset use fluid neutral simulations and no fluid drift effects. The dataset is used to train a simple neural network and Gradient Boosted Regression Trees and test how the performance depends on the number of training simulations.

The influence of the input power (IP), puffing rate and neutral pressure on the fuel (hydrogen isotopes) inventory of the WEST and ITER divertors is investigated. For the chosen range of parameters (relatively low temperature at the strike points), the inventory of the WEST divertor evolves as the power 0.2 of the puffing rate and as the power 0.3 of the IP. The inventory at the strike points is highly dominated by ions whereas it is dominated by neutrals in the private zone. Increasing the fuelling rate increases the retention in the private zone and decreases slightly the retention at the strike points. Increasing the IP increases the inventory at the strike points and does not affect much the inventory at the private flux region. The inventory of the ITER divertor is not strongly dependent on the divertor neutral pressure. The inventory increases from 0Â Pa to 7Â Pa and then decreases slightly from 7Â Pa to 10Â Pa. After 107Â s of continuous exposure, the maximum inventory in the ITER divertor was found to be 14Â g. The inventory is not maximum at the strike points due to the high surface temperature of the monoblocks in this region. The maximum accumulation of H in the ITER divertor is below 5 mg per 400Â s discharge and below 2 mg per 400Â s discharge after 200 discharges. [ABSTRACT FROM AUTHOR]

Comprehensive studies of energy and particle balances in the transition to plasma detachment in an alternative divertor configuration with long outer legs are shown. Numerical simulations are performed with the 2D code suite SOLPS 4.3, using a disconnected double null grid with narrow, tightly baffled long poloidal leg divertors at the outer lower target and outer upper target. A particle count scan is performed using the ‘closed gas box’ model, where the tunable parameter in the simulations is the total number of deuterium particles in the simulation space and all other parameters are held fixed, including a constant input power and trace neon impurity radiation, to assess the physics of the transition to detachment in the system as the particle count increases. Three main aspects of the physics of divertor detachment are addressed: the criteria for the local onset of divertor detachment in each of the divertors, the distribution of heat flux and other plasma parameters between the four divertors as each divertor transitions to detachment, and the role of perpendicular transport in the transition to the detached regime. A synergistic mechanism by which the cross-field transport is reduced by factors associated with the onset of plasma recombination effects is identified. These results are compared to the existing understanding of the physics of the transition to plasma detachment in standard divertors.

SOLPS modelling has shown that divertor plasma detachment occurs at a lower upstream separatrix density in the more closed DIII‐D upper divertor than the open lower divertor, demonstrating the utility of the divertor closure in widening the range of acceptable densities for adequate heat handling. To achieve reduced heat flux and erosion at the plasma‐facing components, future devices will need to operate in at least partially detached divertor conditions . Two‐dimensional fluid plasma models coupled to Monte Carlo neutral transport simulations, such as SOLPS, have been widely used to predict the onset of detachment. In modelling the DIII‐D discharges, the cross‐field transport coefficients are constrained to reproduce the experimental upstream profiles. The closed divertor has been modelled with the same input parameters of the open divertor, allowing a direct comparison of the target conditions in both geometries. SOLPS simulations indicate that a higher molecular density correlates strongly with lower electron temperatures. The increased closure of the upper divertor improves the trapping of neutrals, thereby reducing the power density deposited at the target and facilitating detachment, in agreement with experimental observations. [ABSTRACT FROM AUTHOR]

We employ the SOLPS5.1 code to analyse different impurity choices and injection methods as possible drivers for highly radiative scenarios in the European DEMO (EU‐DEMO). We aim at assessing the existence of a suitable parameter region to safely operate the divertor in H‐mode discharges. It turns out that such an operational region exists, and that puffing is strongly preferred to pellet as the impurity injection method. It also appears that many different impurity mixtures can meet the divertor survival requirements, with a low level of W sputtering. This provides an additional degree of freedom, which will be exploited in the future to optimize the overall reactor performance. [ABSTRACT FROM AUTHOR]