Modelling of a tokamak with reactor technologies was performed with the new SOLPS-ITER 3.2.0 code version, which makes it possible to calculate the plasma parameters up to the main chamber wall. Basing on the modeling, the magnetic equilibrium discussed in the project earlier is compared to the new one presented here for the first time. It is shown that the equilibrium presented earlier can be dangerous for the main chamber wall due to the big electron and ion temperatures in its upper part. The divertor working regimes analyzed earlier using results from synthetic equilibrium and the standard code version SOLPS-ITER 3.0.8, in which the plasma is modeled up to the artificial boundary, are confirmed by more modern modeling described in this work.

This paper presents an overview of recent research advances on the Globus-M2 spherical tokamak over the past two years. Upgrades to the diagnostic complex are reported. A novel inter-shot boronization technique demonstrated superior wall conditioning, achieving effective charge <1.1 and doubling the neutron yield through enhanced fast ion confinement. The hot-ion mode was established as a natural regime, with ion temperatures up to 4.7 keV, revealing that D-neutral beam injection (NBI) provides superior ion thermal insulation compared to H-NBI due to higher toroidal momentum input. Ion cyclotron resonance heating experiments demonstrated effective bulk ion heating and beam ion acceleration. For the first time in a spherical tokamak, the high-field side high-density phenomenon was identified, with SOLPS-ITER modeling confirming the dominant role of plasma drifts. MHD stability studies characterized Alfv & eacute;n eigenmodes causing localized wall loads up to similar to 30 MW/m(2), identified locked modes as a key performance limitation, and demonstrated access to small-edge localized modes (ELM) and ELM-free regimes at reduced triangularity. These results provide the foundation for the design of the next-generation Globus-3 spherical tokamak.

In this work we report the observation of High Field Side High Density Region (HFSHD) in the lower inner divertor of the Globus-M2 tokamak (R = 0.36 m, a = 0.24 m). Discharges with NBI heating (0.7 MW) were investigated using two Thomson scattering diagnostics with probing chords situated at the inner divertor entrance near X-point and in the outer midplane near separatrix. Maximum electron density in the inner divertor leg was found to be 1.5-3 times higher than the density at the outer midplane plasma boundary. The HFSHD region is formed for a wide range of parameters with neat the magnetic axis in the range from 2 x 1019 to 1.4 x 1020m-3. This experimental program was supported by SOLPS-ITER modeling with account of drifts and currents. The neutral particle behavior, including deuterium atoms, molecules and carbon atoms, was modeled with the EIRENE neutral gas transport Monte Carlo code.

The laser-induced quenching diagnostics of atomic hydrogen in the Globus-M2 tokamak divertor is described. The analysis is based on 2D distributions of plasma parameters simulated in SOLPS-ITER code for one of the discharges. Modeling of the expected background signals in spectral lines, as well as the working quenching signals, was performed using a collision-radiative model of deuterium atoms. A time-modulated thulium fiber laser working at 1875 nm with a peak power of 5 W was taken for the modeling as the laser source. The existing Thomson scattering divertor optical layout will be used for both laser probing and collection the fluorescence signal, allowing measurements to be performed in the Globus-M2 divertor inner leg.

The two-dimensional edge plasma code SOLPS-ITER have been used to study the hydrogen isotope effect on the divertor detachment onset for hydrogen (H), deuterium (D), and tritium (T) based on EAST plasma conditions. Results prove that H with a lighter mass has a stronger ability to penetrate into the core plasma than D and T, resulting in a higher density in the core and a lower density in the SOL. If the line-averaged electron density is selected to characterize the detachment onset, the hydrogen isotope effect is strongly dependent on the measurement chord location which determines the contribution of the core electron density to the line-averaged electron density. The modeling results can explain previous hydrogen isotope experimental results in JET. The drifts effect on hydrogen isotope distributions has also been investigated. As the drifts are taken into account, the relative difference of detachment onset at the outer target becomes larger, while it becomes smaller for the inner target. The ExB drifts show a greater impact on the T ion distribution than D and H ions.

The work presents the results of the first calculations of the T-15MD tokamak near-wall plasma in the SOLPS-ITER code taking into account the effects of drifts and currents. The modes with the power passing through the separatrix PSOL = 6 MW and different H gas puff intensities, corresponding to the average electron density on the separatrix, nesep = (2–4.5) × 1019 m–3, are considered. The same as in other tokamaks of similar size, E × B drift leads to the flow of hydrogen from the outer divertor to the inner one, which changes the load distribution between the divertor targets. Drifts also affect the flow of the carbon impurity. As a rule, when describing the dependence on the H gas puff, either nesep or the total amount of hydrogen in the scrape off layer (SOL), Ntot, is used as a parameter characterizing the discharge. In this case, these quantities are considered as equivalent plasma characteristics in the SOL. It is shown that, from the point of view of estimating the influence of drifts, these quantities are not equivalent: the dependence of some divertor parameters on nesep does not change with the inclusion of drifts, but the dependence on Ntot can change. It is also seen that drifts lead to a more pronounced maximum in the dependence of the saturation current on the electron density, Isat(nesep). This is explained by changes in the emission of the carbon impurity and the power of the recombination source of hydrogen in the divertor.

The analysis of possible divertor working regimes and edge plasma parameters for TRT tokamak project is performed basing on modeling. It is shown that for the separatrix power of 18 MW corresponding to approximately twice higher full input power the low divertor integral heat flux 5 MW/m2 can be provided for the separatrix plasma density lower than 7 × 1019 m–3 and the effective charge Zeff lower than 2. These parameters are realistic for this device. In case of bigger separatrix power the working regime is possible with higher divertor heat load still within the technological limits of the machine. Modeling also shows positive effect of the increase of the distance between the separatrix and the vacuum vessel structures and better performance of the corner divertor configuration comparing to the “ITER-like” one.

In the paper a SOLPS-ITER modeling of H-Mode scenarios with different seed impurity gases (N, Ne, Ar, Kr) is performed for ASDEX Upgrade geometry. These gases are compared from the point of view of their radiation efficiency and compression in the divertor, which is generally understood as the ratio of the divertor to upstream density. Three seeding levels of each impurity gas are considered: trace impurity level, level sufficient to radiate half of the incoming power and a so high seeding rate that the X-point radiator (XPR) starts to form. It is demonstrated that Ne has the lowest compression, the compression of Ar is bigger than that of N, and Kr has the best compression among gases considered. Such a ranking corresponds to the ranking of the impurity neutral atom ionization length, which for Kr is the shortest not only due to the smallest first ionization potential, but also due to the biggest ionization cross-section and the biggest mass (or the smallest thermal velocity). Additionally, for the given level of the radiated power fraction frad the seeding of Kr results in the smallest (among other radiants) effective charge Zeff in the confined region, which even for the proximity to XPR does not exceeds 1.5. It appears that the impurity compression strongly depends on whether coronal model or collisional-radiative (CR) one is used for the ionization/recombination rates calculation, or, in other words, whether such processes as step ionization (from excited levels) and 3-body recombination are taken into account. Since for Ar and Kr no results of CR model calculations are available in ADAS database used in SOLPS, the values of compression computed here for these radiants with coronal rates are underestimated. The compression appears to be more sensitive to the neutrals ionization rate than to the details of neutral flow from divertor to the pump, including interaction between neutral species.

For the first time, an X -point radiating (XPR) regime using neon (Ne) seed impurity has been obtained in ITER modelling with the SOLPS-ITER code, including fluid drifts. The highly radiating pattern is qualitatively similar to those observed experimentally in ASDEX Upgrade and JET. Comparison of the simulation results with those obtained from modelling of the same regime, with the same approach, in ASDEX Upgrade is used to analyse the difference between the XPR in large and medium-sized machines. Analysis demonstrates that the cold X -point formation is controlled principally by the Ne radiation. The simulation results for the poloidal scale of the intense radiating zone are in good agreement with analytic estimates.

The processes in far scrape-off layer (SOL) and the plasma interaction with the first-wall (FW) elements may notably affect the tokamak discharge, since they define fuel and impurity recycling, material erosion and redeposition, wall surface heating, etc. For a long time, the far SOL description in most plasma edge transport codes was insufficient or absent at all, and so particle and heat fluxes onto the FW (except divertor plates) were out of consideration. Recently some codes, for example, SOLEDGE and SOLPS-ITER are upgraded allowing for the extension of the computational grid up to real walls and for corresponding account of the vacuum vessel shape and all in-vessel elements. The new release of SOLPS-ITER (the version 3.2.0) required a development of a new code data structure and new approach to numerical approximation of fluid equations compatible with unstructured non-orthogonal computational grid. Intensive testing of the new code in different conditions is still required. In the present contribution, such a testing is performed for the EAST disconnected double null (DDN) L-mode discharge. For the first time, the SOLPS-ITER 3.2.0 modeling results with drifts and currents turned on are presented, and a comparison to former SOLPS-ITER version (3.0.8) is performed. The far SOL transport and its effects on the discharge performance are studied by comparing the computational results obtained on several meshes which differ by their width in equatorial midplane. A single null (SN) one (with mesh width limited by distance to the secondary separatrix) was examined versus two DDN meshes (one with actual and one with artificially extended targets to make mesh wider) and a true unstructured (the widest) mesh. The notable difference in results obtained on different meshes appears in those places where plasma density does not vanishes at the computational domain boundaries. For the cases on true unstructured (the widest) mesh, the particle and heat fluxes onto central column, limiters, far SOL part of targets, dome umbrella and other EAST far SOL in-vessel structures are calculated for the first time by SOLPS-ITER, allowing assessment of the plasma interaction with those surfaces.

The impact of the ionization source in the ITER far SOL associated with main chamber deuterium puffing on the overall SOL performance is studied numerically using SOLPS-ITER modeling. It is found out that this source can have significant influence on the retention of the seeded impurity radiator (Ne) in the divertor and on plasma temperature in far SOL. The former is responsible for typical separatrix concentration of seeded impurity compatible with the required power handling limits on the divertor targets, while the latter is responsible for sputtering of divertor material. Both factors indicate the beneficial effect of high fuel throughput on separatrix and pedestal plasma composition.

SOLPS-ITER modeling of EU-DEMO tokamak burning plasma with Ar seeding was performed. The modeling includes drifts, kinetic neutrals and current description switched on. The simulation results are compared with those without drifts. The power entering the edge plasma domain is 200 MW. The deuterium puff is 1.5 × 1023 at s−1 for all cases, which corresponds to neutral deuterium pressure of 10 Pa in the private flux region. The Ar seeding rates are 8.0×1019  at s−1 and 1.5×1020  at s−1, which correspond to the Ar concentration on the separatrix in the range of 0.5%–2%. It is demonstrated that with such a combination of parameters, it is possible to achieve power loads lower than 5 MW m−2 on both divertor targets. The temperature above 5 eV in the far scrape-off layer (SOL) of the outer target remains an issue. Together with drift and no drift cases, the impurity accumulation mechanism in the high-field-side SOL is discussed.

A new Grad–Zhdanov module is implemented into the SOLPS-ITER code for calculation of the parallel kinetic coefficients. A complete, multi-ion generalization is performed, relaxing the heavy impurity assumption. A JET-like D+T+Ne test simulation is conducted to demonstrate the ability to model a 50/50 deuterium (D) and tritium (T) mixture in SOLPS-ITER. More than 30% T build-up with respect to D is observed in different parts of the simulation domain, in particular at the high field side. A T predominance over D is also observed near both the inner and outer targets. It is a result of the different effective diffusion, dominated by charge-exchange processes, for the D and T neutral species. A simple 1D model is proposed to describe this phenomenon. The contribution to the differing D and T distribution in different regions from the Grad–Zhdanov thermal force is also studied. Due to the thermal force and the D/T poloidal flow from the low field side to the high field side, the prevalence of the T species over D is found at the high field side at the X-point level. The latter leads to an inner–outer divertor asymmetry of n T / n D . However, the effect is relatively small due to the close D and T masses. It is further shown that the infinite ion mass difference limit, which is used for the derivation of the Zhdanov-Yushmanov analytical expressions and applied for the thermal force coefficient calculation previously used in the SOLPS-ITER code, overestimates n T / n D significantly. Thus, the old SOLPS-ITER model should not be applied for D–T simulations. Finally, possible experimental studies of the D and T spatial separation due to the new effects revealed by this modelling are discussed.

Reliable diagnostics that measure the detached state of the ITER divertor plasma will be necessary to control heat flux to the divertor targets during steady state, burning plasma operation. This paper conducts an initial exploration into the feasibility of the divertor shunt diagnostic as a lightweight, robust, and real-time detachment sensor. This diagnostic is a set of shunt lead pairs that measure the voltage drop along the divertor cassette body, from which the plasma scrape-off layer (SOL) current is calculated. Using SOLPS-ITER simulations for control-relevant ITER plasma scenarios, the thermoelectric current magnitude along the SOL is shown to decrease significantly with the onset of partial detachment at the outer divertor target. Electromagnetic modelling of a simplified divertor cassette is used to develop a control-oriented inductance-resistance circuit model, from which SOL currents can be calculated from shunt pair voltage measurements. The sensitivity and frequency-response of the resulting system indicates that the diagnostic will accurately measure SOL thermoelectric currents during ITER operation. These currents will be a good measure of the detached state of the divertor plasma, making the divertor shunt diagnostic a potentially extremely valuable and physically robust sensor for real-time detachment control.

We present the first Thomson scattering (TS) measurements of electron temperature in the lower divertor of the Globus-M2 tokamak. The divertor TS diagnostics is designed for local measurements of the Te(z, t) in the range of 1–100 eV and ne(z, t) in the range of ∼ 1pt10^17-10^20 m–3. Parameters of the probing Nd:YAG laser are as follows 1064 nm/2 J/100 Hz/3 ns. The probing chord is launched vertically at R = 24 cm and covers areas of the inner leg, vicinity of separatrix and private flux region. Along probing chord of 110 mm, 9 spatial points were realized. Advanced filter polychromators were used to analyze Thomson scattering spectra.

In this paper SOLPS-ITER simulations of Ar and N seeded discharges are presented for H-mode conditions on ASDEX Upgrade and for burning plasmas on ITER. These discharges are additionally compared with Ne seeded cases for both tokamaks. It is demonstrated that Ar and N are retained in the divertor, and that the amount of seeded impurity (measured both by the seeding rate in atoms/s and the averaged separatrix density) necessary to achieve the same divertor conditions is lower for Ar than for N. This is attributed to almost the same values of first ionization potential of Ar and N (if compared to e.g. Ne) and to higher Ar radiation efficiency. If the radiated power fraction within the numerical simulation grid is of the order of 50%, which is sufficient to achieve partial detachment at the outer target in both devices, radiation losses from the confined region remain small for Ar seeded ITER cases (about 6% of power from the outer core region included in the SOLPS-ITER computational domain with total volume 135 m3). The fuel dilution by Ar remains lower than that by N and Ne, which corresponds to an Ar averaged separatrix concentration of 0.4%. From the point of view of target power handling, the boundary simulations thus demonstrate that Ar may be a suitable alternative seed impurity on ITER.