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An H-mode database of ASDEX Upgrade plasmas with improved diagnostics for a variety of different plasma conditions has been analysed to study the momentum removal in the Scrape-Off Layer (SOL). A strong reduction, up to a factor of 100, of the electron pressure is observed close to the separatrix. Sets of L-mode and H-mode like plasma simulations with the SOLPS5.0 code package have supported the interpretation of the experimental observations (e.g., apparent momentum gain, strong radial position dependence) and helped to clarify the role of ion-neutral interactions. The experimental data are in turn used to check the SOLPS predictions, which differ quantitatively, suggesting that important pieces of physics are still not successfully captured by the code. Keywords: Momentum removal, Scrape-off layer, SOLPS, Divertor detachment

The experimentally observed in/out detachment asymmetry in KSTAR L-mode plasmas with deuterium (D) fueling and carbon walls has been investigated with the SOLPS-ITER code to understand its mechanism and identify important atomic processes in the divertor region. The simulations show that the geometrical combination of a vertical, inner target with a short poloidal connection from the X-point to the target and a much longer outer divertor leg on an inclined target lead to neutral accumulation towards the outer target, driving the outer target detachment at lower upstream density than is required for the inner target. This is consistent with available Langmuir probe measurements at both target plates, although the inner target profile is poorly resolved in these plasmas and further experiments with corroborating diagnostics are required to confirm this finding. The pressure and power loss factors defined in the two-point model (Stangeby 2018 Plasma Phys. Control. Fusion 60 4; Kotov and Reiter 2009 Plasma Phys. Control. Fusion 51 115002; Stangeby and Sang 2017 Nucl. Fusion 57 056007; Moulton et al 2017 Plasma Phys. Control. Fusion 59 6) of the divertor scrape-off layer (SOL) and the sources contributing to the loss factors are calculated through post-processing of the SOLPS-ITER results. The momentum losses are mainly driven by plasma-neutral interaction and the power losses by plasma-neutral interaction and carbon radiation. The presence of carbon impurities in the simulation enhances the pressure and power dissipation compared to the pure D case. Carbon radiation is a strong power loss channel which cools the plasma, but its effect on the pressure balance is indirect. Reduction of the electron temperature indirectly increases the momentum loss and increasing the volumetric reaction rates which are responsible for the loss of momentum. As a result, the addition of carbon saturates the momentum and power losses in the flux tube at lower upstream densities, reducing the roll-over threshold of the upstream density. The relative strengths of the various mechanisms contributing to momentum and power loss depend on the radial distance of the SOL flux tubes from the separatrix (near/far SOL) and the target (inner/outer target). This is related to the strong D2 molecule accumulation near the outer strike point, which makes the deuterium gas density at the outer target 2-10 times higher than that at the inner target. A large portion of the recycled neutral particles from both targets reach and accumulate in the outer SOL, which is predominantly attributed to the target inclination and gap structure between the central and outboard divertors and hence to the impact of geometry. The accumulated neutrals enhance the reactions involving D2, which causes momentum and power loss.

© 2018 IAEA, Vienna. The baseline ITER burning plasma equilibrium is designed to place the divertor strike points deep into the 'V-shaped' region formed by the high heat flux handling vertical targets (VT) and the reflector plates (RP). The divertor plasma performance under these conditions has been extensively studied in the past two decades with the SOLPS4.3 plasma boundary code suite. However, during tokamak operation, inaccuracies in the control of the vertical plasma position, or a requirement to avoid damaged monoblocks, could force the strike point position further down the VTs, or even directly on the RPs. In this paper, we present the results from the first SOLPS-ITER modelling in which the consequences of strike point displacements on the divertor plasma behaviour and surface heat loading are assessed. The starting point of the study is a baseline coupled fluid plasma-kinetic neutral solution (without fluid drifts), corresponding to an ITER burning plasma scenario at Q DT = 10 with neon seeding for detachment control, P SOL = 100 MW, λ q ∼ 2 mm and nominal strike point positions. From this baseline condition, the equilibrium is progressively moved downwards in a series of rigid displacements, obtaining new steady-state solutions, up to a maximum displacement of ∼8 cm, beyond which the separatrix is too close to the inner dome wing. At this point, the inner strike point is well onto the inner RP while the outer strike point is still on the VT. The different interaction of the recycled neutrals with the SOL plasma when the strike point intersects the inner RP, switching from vertical to horizontal target configuration, enhances the detachment degree at the inboard divertor, mitigating the heat load deposited onto the inner RP. At the outboard divertor the plasma condition is not significantly affected by the downward displacements, nor are the power fluxes to the outer RP. Finally, the heat load profiles computed with SOLPS are used in input for a finite element thermal analysis, considering the full cooling geometry, to assess the response of the VTs and RPs under the conditions exploited in the displaced scenarios. This thermal model, based on a simplified treatment not requiring a full 3D description of the divertor monoblock plasma-facing units, constitutes a new module for the SOLPS-ITER code suite. ispartof: NUCLEAR FUSION vol:58 issue:12 status: published

V simulacijah plasti postrgane plazme (SOL) z B2-EIRENE oziroma SOLPS (Scrape-Off Layer Plasma Simulations) se pojavlja veliko število vhodnih parametrov, možnih kombinacij ionskih delcev, atomskih procesov in interakcij med plazmo in steno. Simulacije SOLPS se izvajajo na superračunalnikih v uporabniških imenikih in se lahko shranjujejo tudi v podatkovne baze MDSPLUS ali pretvorijo za zapis v fuzijske podatkovne baze s predpisanimi strukturami za integrirano modeliranje (EUROfusion CPO ali ITER IDS). To magistrsko delo raziskuje podatkovno strukturo EUROfusion CPO in ITER IDS ter pridobljene informacije uporabi za pripravo primernih orodij za transformacijo podatkov iz CPO v IDS ter izdelavo vtičnika in nadgradenj splošno-namenskega vizualizacijskega orodja ParaView, ki bo omogočal primerjalne analize rezultatov simulacij SOLPS. Nenazadnje je bila nadgrajena koda SOLPS-ITER, ki omogoča zapis edge_profiles IDS-jev v podatkovno bazo IMAS in tako nudi nabor orodij za okolje integriranega modeliranja. In Scrape-Off Layer Plasma Simulations (SOLPS) with B2-EIRENE there are many input parameters, possible combinations of ion particles, atomic processes, and interactions between the edge plasma and the core boundary and plasma-facing components. The SOLPS are carried out on a cluster in a user directory and can be stored in the MDSPLUS database or can be converted to a "record" in fusion databases with prescribed structures for integrated modeling (EUROfusion CPO or ITER IDS). This thesis investigates the data structure of EUROfusion CPO and ITER IDS and uses gathered information to create appropriate tools for data transformation from CPO to IDS together with plugins and upgrades for general purpose visualization tool ParaView which allows comparative analysis of SOLPS output results. Finally, the SOLPS-ITER code is upgraded to write edge_profiles IDS into IMAS database providing a required set of tools for in integrated modeling framework.

In this work, an investigation of the neutral gas flow in the JET sub-divertor area is presented, with respect to the interaction between the plasma side and the pumping side. The edge plasma side is simulated with the SOLPS code, while the sub-divertor area is modeled by means of the direct simulation Monte Carlo (DSMC) method, which in the last few years has proved well able to describe rarefied, collisional flows in tokamak sub-divertor structures. Four different plasma scenarios have been selected, and for each of them a user-defined, iterative procedure between SOLPS and DSMC has been established, using the neutral flux as the key communication term between the two codes. The goal is to understand and quantify the mutual influence between the two regions in a self-consistent manner, that is to say, how the particle exhaust pumping system controls the upstream plasma conditions. Parametric studies of the flow conditions in the sub-divertor, including additional flow outlets and variations of the cryopump capture coefficient, have been performed as well, in order to understand their overall impact on the flow field. The DSMC analyses resulted in the calculation of both the macroscopic quantities—i.e. temperature, number density and pressure—and the recirculation fluxes towards the plasma chamber. The consistent values for the recirculation rates were found to be smaller than those according to the initial standard assumption made by SOLPS.

Numerical simulation of the tokamak scrape-off layer (SOL) is an essential tool for the prediction of the conditions to be expected in future fusion reactors such as the ITER project, now under construction in Southern France. One particularly important issue regards the estimation of the expected transient power loads on plasma-facing components (PFC) due to magnetohydrodynamic plasma relaxations, known as Edge Localised Modes (ELMs). These loads are a major cause of concern for ITER owing to the very severe restrictions on PFC lifetime (especially the divertor targets) that they will impose if their amplitude is not maintained below a given size. Even though SOL plasma modelling has reached a comparatively high level of sophistication (the ITER divertor is being designed in part with complex edge plasma codes), the majority of simulations are performed for steady state conditions, necessarily excluding the description of transient events. This thesis explores the utility and validity of the fluid plasma, Monte-Carlo neutrals approach to SOL and divertor modelling in the presence of time dependent ELM phenomena. It aims to test the most complex tool of this type currently available, the fluid (B2.5)-neutral Monte-Carlo (EIRENE) code package SOLPS5, against a variety of ELM sizes in two very different tokamaks, TCV and JET. Although the SOLPS package has been the modelling tool of choice for ITER design, it has not yet been systematically used for the study of ELM transients. A key element throughout is rigorous benchmarking – seeking the best possible agreement between both experiment and simulation and between different codes for the same experiment, using as many different measurements as possible to constrain the model. Such benchmarking attempts are still, unfortunately, comparatively rare on today's machines. Fully time-dependent simulations (2-D plasma, 3-D neutrals) have been performed of four H-mode plasmas, two each on TCV and JET, covering Type III and Type I ELMs over a range of pedestal collisionality and energy expelled per ELM from ΔWELM ∼ 0.005 → 0.7 MJ. The high end of this limit corresponds to the current maximum ΔWELM which is thought to be tolerable on ITER for acceptable divertor target lifetime. The two tokamaks differ radically in size, input power and divertor geometry, but share carbon as the main PFC material. The SOLPS5 simulations have thus been performed with all carbon charge states included but do not feature activated poloidal drift terms. The approach is first to seek the closest match to experimental upstream, pedestal/SOL and downstream target profiles during the inter-ELM phase. This is achieved through the specification of radially varying perpendicular particle and heat diffusivities and/or convective radial velocity in order to account for the different transport levels in the edge and SOL regions. Poloidal variation of these transport coefficients is also applied to distinguish between main chamber SOL and divertor regions. This is important in a device like TCV with rather unconventional divertor geometry. Similar reasoning applies even more to the ELM itself, which is known to burst into the SOL in the outboard, unfavourable curvature region and is thus extremely poloidally localized. This has also been accounted for especially in the attempts to simulate the TCV ELM events. The complexity of the ELM instability continues to prevent a complete theoretical description of the evolution of transport during the event. In SOLPS5, the simplest and currently only method by which the ELM can be simulated is to increase the anomalous transport coefficients used to simulate the pre-ELM state during a brief interval corresponding to the ELM duration, such that the total energy expelled during this time is compatible with that measured experimentally. In the case of TCV Type III ELM, where reasonable upstream and downstream data are available and for which the largest number of sensitivity studies have been performed in this thesis with SOLPS, agreement is good in the pre-ELM phase and reasonable, but less satisfactory during the ELM. This ELM is a largely convective event in terms of pedestal temperature collapse and is, by virtue of its low ΔWELM, the "least kinetic" of the four events studied. Nevertheless, comparison of the SOLPS5 simulation results at the divertor target with those from dedicated Particle-in-Cell kinetic transport code calculations for the same ELM, demonstrate that kinetic effects are important and must be properly accounted for (by appropriate adjustment of kinetic coefficients in the fluid simulations). This presumably becomes even more important as the ELM size increases, but can only be tested to the extent that the appropriate experimental data is available. As a consequence, the tentative conclusion from the work presented here is that the use of SOLPS in a predictive sense for ITER would at best provide indicative results. In addition to the code-experiment benchmark, a code-code comparison has also been performed, checking SOLPS5 against published and well known time dependent Type I ELM simulations obtained with the dedicated JET code suite EDGE2D-Nimbus. A benchmark of this complexity has not previously been attempted and has been reassuringly somewhat successful, albeit with some unresolved discrepancies. A key feature of ELM boundary physics occupying much current research are the energy deposition asymmetries observed at the targets, which favour the inner target during the ELM for forward toroidal field direction and which appear to reverse when the field direction is inverted. These trends are opposite to the behaviour seen in inter-ELM phases, behaviour which is conventionally understood to result from toroidal geometry and the contribution of poloidal drift physics. Added complexity comes from magnetic geometry, a prominent feature of the TCV-JET comparisons described in this thesis, the results of which seem to influence the simulation results (which do not include drift effects). A recent development has been the suggestion that the ELM, in convecting plasma from pedestal to SOL regions, carries with it memory of the high toroidal rotation velocity known to characterise the H-mode pedestal on all devices. This hypothesis has been tested here in a preliminary manner, and for the first time in this kind of simulation, by imposing a toroidal velocity inside the magnetic separatrix in the simulations and studying the radial transport of this toroidal momentum into the SOL. Applied in the first instance to the TCV Type III ELM, the indications are that transfer of this rotation into the SOL can drive target asymmetries in the direction seen experimentally, though there are significant negative consequences for the resulting target profiles in other parameters for which a potential resolution would require protracted further study which has not been possible here.

In this paper we present a computational study on the divertor heat load mitigation through impurity injection for the EU DEMO. The study is performed by means of the SOLPS5.1 code. The power crossing the separatrix is considered fixed and corresponding to H-mode operation, whereas the machine operating condition is defined by the outboard mid-plane upstream electron density and the impurity level. The selected impurity for this study is Ar, based on its high radiation efficiency at SOL characteristic temperatures. We consider a conventional vertical target geometry for the EU DEMO and monitor target conditions for different operational points, considering as acceptability criteria the target electron temperature (≤5 eV to provide sufficiently low W sputtering rate) and the peak heat flux (below 5-10 MW m-2 to guarantee safe steady-state cooling conditions). Our simulations suggest that, neglecting the radiated power deposition on the plate, it is possible to satisfy the desired constraints. However, this requires an upstream density of the order of at least 50% of the Greenwald limit and a sufficiently high argon fraction. Furthermore, if the radiated power deposition is taken into account, the peak heat flux on the outer plate could not be reduced below 15 MW m-2 in these simulations. As these simulations do not take into account neutron loading, they strongly indicate that the vertical target divertor solution with a radiative front distributed along the divertor leg has a very marginal operational space in an EU DEMO sized reactor. © 2018 Politecnico di Torino.