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Данная бакалаврская работа посвящена модификации программы solps-iter для учёта нелинейного теплообмена между электронами и ионами в пристеночной плазме токамаков. В работе проведен обзор Ñ‡Ð¸ÑÐ»ÐµÐ½Ð½Ñ‹Ñ Ð¼ÐµÑ‚Ð¾Ð´Ð¾Ð² решения Ð½ÐµÐ»Ð¸Ð½ÐµÐ¹Ð½Ñ‹Ñ ÑƒÑ€Ð°Ð²Ð½ÐµÐ½Ð¸Ð¹, описание внутренней реализации программы solps-iter, ее модификация для учета нелинейного теплообмена между электронами и ионами. Также произведены тесты, сравнивающие скорость ÑÑ Ð¾Ð´Ð¸Ð¼Ð¾ÑÑ‚Ð¸ старого и нового варианта программы для Ñ€Ð°Ð·Ð»Ð¸Ñ‡Ð½Ñ‹Ñ Ñ‚Ð¾ÐºÐ°Ð¼Ð°ÐºÐ¾Ð². Результатом работы является модернизированная программа solps-iter, которая с точки зрения математики верно учитывает нелинейный теплообмен между электронами и ионами и решает уравнения плазмы со ÑÑ Ð¾Ð´Ð¸Ð¼Ð¾ÑÑ‚ÑŒÑŽ не Ñ ÑƒÐ¶Ðµ, чем у предыдущего варианта программы., This bachelor's work is devoted to the modification of the solps-iter program to account for the nonlinear heat transfer between electrons and ions in the wall plasma of tokamaks. The paper contains: review of numerical methods for solving nonlinear equations, description of the internal implementation of the solps-iter program, its modification to account for the nonlinear heat transfer between electrons and ions. It also contains the results of tests comparing the convergence rate of the old and new versions of the program for different tokamaks. The result is a modernized solps-iter program, which, from the point of view of mathematics, correctly takes into account the nonlinear heat transfer between electrons and ions and solves the plasma equations with convergence no worse than that of the previous version of the program.

Linear plasma devices represent an essential tool for nuclear fusion research, whereby understanding crucial aspects related to plasma-wall interactions or edge plasma behaviour. Simplified models are of great importance to complement and integrate experimental and simulation results of complex systems such as plasmas in linear machines, because they are fast and simple to employ. In this work, we present a global volume-averaged (0D) model for plasma investigation in linear machines. The 0D model equations are based on the space integration of the state of the art edge plasma model implemented in the SOLPS-ITER code. Comparisons between helium plasmas described with 2D simulations performed with SOLPS-ITER and with the 0D model highlight that contributions often neglected in tokamak edge models, e.g. electron-neutral excitation, may be relevant when describing weakly ionized plasmas in linear devices. The model is used to perform sensitivity studies with respect to several parameters and to analyse the time evolution of the system, leading to the identification of two relevant time scales governing the system. Lastly, a comparison of 0D results with experimental data from the linear device GyM is performed, showing satisfactory agreement. Our methods and results provide crucial interpretative keys in the investigation of the physics of edge plasmas.

A double-null configuration is being considered for the EU-DEMO, due to its potential benefits for power exhaust arising from the use of two active divertors and magnetically disconnected low- and high-field sides. Using systematic parameter scans in fluid simulations, we have investigated the divertor power exhaust in the EU-DEMO in a connected double-null configuration, and compared the edge plasma properties to those obtained in a single-null configuration under detached conditions anticipated for reactor operation. Neglecting drift effects and kinetic behaviour of the neutrals, no clear benefits of the double-null configuration could yet be identified for the radiation pattern and power mitigation on open field lines. Future work should address the aforementioned physics as well as the effect of the additional X-point on core radiation.

Energy dissipation in the plasma edge is key for future tokamaks. The potential of neon as radiating seeding species in disconnected double null (DDN) configuration is assessed in EAST discharges in high confinement mode (H-mode). As the separation between the two separatrices in the studied DDN discharges is minimum 1.5 cm, the configuration is effectively a single null configuration, and the benefits of the double null topology are minimal. Neon seeding, on the other hand, has a favourable effect: both the target heat flux and the divertor temperature decrease more than five-fold with increased seeding rate in high-recycling conditions. Interpretive edge plasma simulations with SOLPS-ITER in support of ongoing transport analysis are presented. For the unseeded case the numerical results agree with the experimental data within a factor two for the target temperature conditions and measured neutral pressures in the active divertor. The key for achieving good agreement is a suitable selection of coefficients for anomalous transport and neutral conductances between the upper cryopump and the main chamber.

Данная работа посвящена моделированию кодом SOLPS-ITER разряда компактного сферического токамака Глобус-М2 на основе ÑÐºÑÐ¿ÐµÑ€Ð¸Ð¼ÐµÐ½Ñ‚Ð°Ð»ÑŒÐ½Ñ‹Ñ Ð´Ð°Ð½Ð½Ñ‹Ñ , Ð¿Ð¾Ð»ÑƒÑ‡ÐµÐ½Ð½Ñ‹Ñ Ð¿Ñ€Ð¸ помощи диагностики Томпсоновского рассеяния лазерного излучения, а также анализу влияния примеси на разряд на основе моделирования и изучению возможности Ð¿ÐµÑ€ÐµÑ Ð¾Ð´Ð° в режим отрыва Ð´Ð¸Ð²ÐµÑ€Ñ‚Ð¾Ñ€Ð½Ñ‹Ñ Ð¿Ð»Ð°ÑÑ‚Ð¸Ð½. Проблема снижения потоков тепла на диверторные пластины стоит наиболее остро на ÑƒÑÑ‚Ð°Ð½Ð¾Ð²ÐºÐ°Ñ Ñ высокой мощностью и сравнительно длительным временем удержания, так как на Ñ‚Ð°ÐºÐ¸Ñ ÑƒÑÑ‚Ð°Ð½Ð¾Ð²ÐºÐ°Ñ Ð¿Ð»Ð¾Ñ‚Ð½Ð¾ÑÑ‚ÑŒ потоков мощности на диверторные пластины не может превышать критическое значение 10 МВтм2 . И Ñ Ð¾Ñ‚ÑŒ в токамаке Глобус-М2 данная проблема не стоит остро, он является отличной площадкой для изучения физики управляемого термоядерного синтеза. Основной идеей Ð¿ÐµÑ€ÐµÑ Ð¾Ð´Ð° в режим отрыва Ð´Ð¸Ð²ÐµÑ€Ñ‚Ð¾Ñ€Ð½Ñ‹Ñ Ð¿Ð»Ð°ÑÑ‚Ð¸Ð½ при помощи напуска излучающей примеси является сокращение потока мощности на диверторные пластины посредством излучения напускаемой примеси. При этом стоит учитывать распределение примеси в установке и не допускать ее скопления в зоне удержания. Иными словами, важно так подобрать примесь, скорость напуска и место напуска, чтобы подавляющий процент излученной мощности Ð¿Ñ€Ð¸Ñ Ð¾Ð´Ð¸Ð»ÑÑ на регионы, не являющиеся зоной удержания. В данной работе приведены результаты моделирования разряда токамака Глобус-М2 с различной скоростью напуска азота. Случай с минимальным напуском азота Ñ Ð¾Ñ€Ð¾ÑˆÐ¾ согласуется с экспериментальными данными, полученными в Ñ Ð¾Ð´Ðµ разряда №38361 на 197 мс. Из результатов моделирования с увеличением скорости напуска азота можно сделать вывод, что Ð¿ÐµÑ€ÐµÑ Ð¾Ð´ в режим отрыва возможен на рассматриваемой установке в режиме H-mode, что совпадает с экспериментами, проводимыми на установке ASDEX-Upgrade[10]. Однако в силу отстутствия экспериментального подтверждения остается открытым вопрос не повторит ли Глобус-М2 сценарий разряда в токамаке COMPASS[8], на котором удалось получить лишь режим отрыва в L-mode, как отмечалось, в силу недостаточности мощности дополнительного нагрева и Ð²Ñ‹ÑÐ¾ÐºÐ¸Ñ Ð¿Ð¾Ñ‚ÐµÑ€ÑŒ на излучение в зоне удержания., This work is devoted to the reduction of heat fluxes to the divertor and to the studying of the possibility of using nitrogen as a radiating impurity for switching to the detachment mode on the compact spherical tokamak Globus-M2 by modeling with SOLPS-ITER code based on experimental data of Thomson scattering diagnostics. Reduction of heat flows to divertor plates is one of the most urgent problem in devices with high power and a relatively long pulse duration since in such devices the density of power flows to divertor plates should not exceed the critical value 10 МWtm2 . Although this problem is not acute in the Globus-M2 tokamak, it is an excellent platform for studying the physics of controlled nuclear fusion. The main idea of switching to the detachment mode by the radiating impurity seeding is reducing the power flow to the divertor plates due to radiation loss by the input impurity. At the same time, it is necessary to take into account the distribution of the impurity in the device and avoid its accumulation in the confinement zone. Also, it is important to choose the type of impurity, the impurity injection rate and the location so that the overwhelming percentage of the radiated power falls on regions which are not the confinement zone. This work presents the results of modeling the discharge of the Globus-M2 tokamak with different nitrogen injection rates. The purest of the presented cases agrees well with the experimental data obtained during discharge №38361 at 197 ms. According the modeling results with an increase of the nitrogen injection rate, it can be concluded that the transition to the detachment is possible in H-mode at the device , which coincides with the experiments conducted at the ASDEX-Upgrade tokamak[10]. However, due to the lack of experimental data the question remains open whether the Globus-M2 will repeat the discharge scenario of the COMPASS[8] tokamak where it was possible to obtain only detachment in L-mode, as noted, due to insufficient additional heating power and high radiation losses in the confinement zone.

Данная работа посвящена моделированию Ð´Ð¸Ð½Ð°Ð¼Ð¸Ñ‡ÐµÑÐºÐ¸Ñ Ñ€ÐµÐ¶Ð¸Ð¼Ð¾Ð² работы токамака кодом SOLPS-ITER. Исследование Ñ‚Ð°ÐºÐ¸Ñ Ñ€ÐµÐ¶Ð¸Ð¼Ð¾Ð² Ð½ÐµÐ¾Ð±Ñ Ð¾Ð´Ð¸Ð¼Ð¾, поскольку в Ñ Ð¾Ð´Ðµ Ð½ÐµÑÑ‚Ð°Ñ†Ð¸Ð¾Ð½Ð°Ñ€Ð½Ñ‹Ñ Ñ€Ð°Ð·Ñ€ÑÐ´Ð¾Ð² конфигурация плазмы претерпевает существенные изменения, а такие Ñ Ð°Ñ€Ð°ÐºÑ‚ÐµÑ€Ð¸ÑÑ‚Ð¸ÐºÐ¸ как температура и потоки тепла, Ð¿Ñ€Ð¸Ñ Ð¾Ð´ÑÑ‰Ð¸Ðµ на материальные Ð¿Ð¾Ð²ÐµÑ€Ñ Ð½Ð¾ÑÑ‚Ð¸ камеры, Ð½ÐµÐ¾Ð±Ñ Ð¾Ð´Ð¸Ð¼Ð¾ отслеживать на протяжении всего разряда, даже если последний Ð¿ÐµÑ€ÐµÑ Ð¾Ð´Ð¸Ñ‚ к стационарному режиму. В данной работе отдельно исследуется класс решений, который носит колебательный Ñ Ð°Ñ€Ð°ÐºÑ‚ÐµÑ€. Для него правильное динамическое описание является единственно возможным. Такой режим работы токамака наблюдался в ÑÐºÑÐ¿ÐµÑ€Ð¸Ð¼ÐµÐ½Ñ‚Ð°Ñ Ð½Ð° Ñ‚Ð¾ÐºÐ°Ð¼Ð°ÐºÐ°Ñ ASDEX Upgrade и JET. Моделирование проводилось версией кода 3.0.6, которая ранее использовалась для поиска стационарной конфигурации пристеночной плазмы токамака. Тем не менее, численная ÑÑ ÐµÐ¼Ð° в нем реализована так, что на каждой итерации ищется приращение Ð¸ÑÐºÐ¾Ð¼Ñ‹Ñ Ð²ÐµÐ»Ð¸Ñ‡Ð¸Ð½ по времени, то есть код в принципе должен позволять рассчитывать динамические изменения, при условии достаточно малого шага по времени. В серии Ñ‚ÐµÑÑ‚Ð¾Ð²Ñ‹Ñ Ð´Ð¸Ð½Ð°Ð¼Ð¸Ñ‡ÐµÑÐºÐ¸Ñ Ñ€Ð°ÑÑ‡ÐµÑ‚Ð¾Ð², Ð¿Ñ€Ð¾Ð²ÐµÐ´ÐµÐ½Ð½Ñ‹Ñ Ð² данной работе, в геометрии токамака CFETR были выявлены решения колебательного типа. При более детальном анализе профилей параметров плазмы было обнаружено, что численная ÑÑ ÐµÐ¼Ð° реализованная в версии кода SOLPS-ITER 3.0.6, дающая физически правильные стационарные решения, приводит к численно неверным результатам при попытке применить ее к динамической задаче., This work is devoted to model of dynamic operation modes of a tokamak using the SOLPS-ITER code. The study of such modes is necessary, since during non- stationary discharges the configuration of a plasma undergoes significant changes. Magnitudes of such characteristics as a temperature and heat fluxes arriving at the material surfaces of the tokamak chamber must be monitored throughout the entire discharge, even if the latter tends to a stationary one. In this work we specially investigate the class of solutions, which has an oscillatory nature. In such case a correct dynamic description is the only possible one. Such operation regime was previously observed in experiments on the ASDEX Upgrade and JET tokamaks. The simulation was carried out with code version 3.0.6, which was previously used to search for the stationary configuration of the plasma. Nevertheless, the numerical scheme in this code at each iteration calculate the time variation of the sought-for values. So, in principle, this version of the code can calculate dynamic changes, if the time step is sufficiently small. In a series of the test simulations carried out in this work, oscillations were revealed in the geometry of the CFETR tokamak. Detailed analysis of the profiles of plasma parameters revealed that the numerical scheme implemented in version 3.0.6 of the SOLPS-ITER code, which gives physically correct stationary solutions, leads to numerically incorrect results when we try to apply it to a dynamic problem.

Systematic extraction of locally valid dynamic models from experiments is necessary for controller design and the validation of high fidelity models. This paper describes the extraction of a dynamic model in the form of a transfer function, giving the dynamic response of the CIII (465.0 nm) emission front position to deuterium gas puffing in the TCV divertor during flattop, relevant for heat exhaust control. The model is extracted using frequency response data from both SOLPS-ITER simulations and perturbative experiments. We use the steady-state solutions of the model SOLPS-ITER to obtain an additional data point at the zero frequency, as the identifiable frequency range by perturbative experiments is lower bounded by discharge time. We specifically approach the problem from a control engineering point of view, aiming to develop control-oriented models for the systematic design of impurity emission front controllers. We find a transfer function structure based on a diffusive process to best describe the obtained frequency response data. The resulting transfer function model accurately reproduces the local dynamic response measured during experiments, so it can be used to assess new controllers offline for similar discharge scenarios.

Edge plasma codes, such as SOLPS, are widely used to study plasma transport in tokamaks scrape-off layers (SOL). The possibility to apply these codes to non hydrogenic plasmas and to linear plasma devices is gaining the interests of the fusion community. These facilities play a pivotal role in plasma-material interaction studies for future fusion devices and may allow to test the code capabilities both in terms of geometry and simulated plasma species. In this contribution, we apply the SOLPS-ITER code for the simulations of Argon plasmas in the medium-flux linear machine GyM. A sensitivity scan over the pumping efficiency, transport coefficients and absorbed electron power was done, performing B2.5-EIRENE coupled simulations. A quantitative analysis of the different flux contributions is provided through the use of a two-point modelling. Comparison with experiments shows a promising qualitative and quantitative agreement, with the sole exception of the simulated neutrals pressure. Dedicated EIRENE standalone simulations were performed to investigate this issue, also highlighting the role of neutral-neutral elastic collisions at high values of the puffing intensity.

A numerical study is presented, using the SOLPS-ITER plasma boundary code with full drifts and currents activated, of impurity seeded JET discharges in support of high power H-mode experimental campaigns designed to compare two ITER candidate seeding species, nitrogen (N) and neon (Ne). Fluid-kinetic edge plasma simulations are first performed at lower levels of power into the scrape-off layer (SOL) and benchmarked against existing JET experimental data. Calculations are then performed for higher levels of SOL power to examine the impact of this key parameter on the efficiency of both radiators. From the code point of view, for the chosen JET parameters, Ne can be as efficient a divertor radiator as N at the same level of upstream separatrix Zeff

The SOLPS-ITER code suite is used worldwide for plasma edge modeling, the interpretation of experiments, as well as for the design of the ITER divertor. The numerical scheme of the plasma solver of the code, B2.5, is based on the assumption of perfectly orthogonal grids in the poloidal plane, aligned with the magnetic field, while in practice grids are often strongly distorted to match divertor target shapes. Neglecting these grid distortion leads to qualitatively and quantitatively incorrect results for fluid neutral simulations, and may affect results in cold (detached) divertors even when using kinetic neutral simulations. In this contribution, we present the first results of a newly implemented 9-point stencil in B2.5 to properly handle misaligned grids. The new scheme is then applied to fluid neutral simulations of a well-diagnosed and previously modeled Alcator C-Mod discharge. Results are compared with the original 5-point scheme neglecting grid distortion effects, as well as with simulations including a full kinetic neutral model. We conclude that the 9-point stencil is essential to correctly model the transport of fluid neutrals on distorted grids, and to capture the effects of divertor closure on the fluid neutral behavior. Keywords: SOLPS-ITER, Divertor modeling, Fluid neutral modeling, Misaligned grids, Alcator C-Mod

In the context of magnetic confinement fusion research, understanding and controlling plasma-material interaction (PMI) is a key point for the realisation of future fusion device, such as ITER. Since the conditions foreseen in ITER have never been obtained before, there is the need to study PMI in ITER-relevant regimes both numerically and experimentally. Linear plasma devices are widely adopted as material-testing facilities, being able to generate ITERrelevant plasmas in a compact and cost-effective fashion. On the numerical aspect, dedicated codes have been developed by the fusion community addressing the modelling of edge plasmas (e.g. SOLPS [1]) and plasma material interaction (e.g. ERO [2]). Edge plasma codes are widely used for the modelling of present-day tokamak devices, but are scarcely applied to linear machines. To bridge this gap, in this contribution, we show the first results concerning the application of the SOLPS-ITER [3] code to the medium-flux linear plasma device GyM[4] (figure 1). As a starting point, simplified argon plasmas were simulated, considering only one charged state (Ar+) and the neutral atom. The atomic nature of the Ar gas offers a simplified picture to apply the code, in absence of the complexity related to molecular species. A detailed sensitivity scan on several code free parameters (such as puffing, pumping efficiency, absorbed power, etc.) is first performed. Simulations results are then compared to experimentally available data, showing a remarkably good quantitative and qualitative agreement between numerical simulations and experimental data Figure 1: Schematic drawing of the linear plasma device GyM.

Данная работа посвящена моделированию токамака-реактора ИТЭРс примесью бериллия кодом SOLPS-ITER. В процессе работы изучен перенос бериллия в Ñ€Ð°Ð·Ñ€ÑÐ´Ð°Ñ , которые ожидаются в термоядерной кампании ИТЭРа. Были выявлены основные факторы, определяющие осаждение. Проведено сравнение с результатами кода WallDYN., This tesis is devoted to modeling the ITER tokamak reactor with beryllium impurity using the SOLPS-ITER code. In is paper the transport of beryllium in the discharges expected in the ITER thermonuclear campaign was studied. The main factors determining deposition were identified. A comparison is made with the results of the WallDYN code.

To understand the beneficial effects of impurity boron (B) on the plasma, the transport of B species in both atomic and ionic states in the edge plasma and the ensuing evolution of the background plasma are simulated during real-time B powder injection by using two-dimensional fluid code SOLPS-ITER. The initial background plasma profiles for the simulation are reconstructed based on an EAST upper-single null discharge (shot No. 70601). With the injection rate of B atoms, 4.0 × 1021 atoms/s, the simulation shows that effective charge number is less than 1.1 at the inner boundary R - Rsep = −5.2 cm (background plasma density, 5.1 × 1019 m−3) at the outer mid-plane indicating that the B impurity does not have much influence on the energy confinement. The present work finds that the impurity B exhibits very specific features in the edge plasma, mostly present in the narrow region just outside the separatrix. The particular distribution pattern can reduce the heat flux greatly, leading to the divertor detachment. The divertor detachment can be achieved easily under an injection rate of B powder in a reasonable range. This study suggests that the impurity B has a promising potential application in future tokamaks.

EAST was equipped with W-coated divertor plates and dome for the upper divertor, the remaining plasma facing components include the plasma facing components at lower divertor consisted of Carbon materials, at the baffles and main chamber walls consisted of Molybdenum materials. In order to study the control of W impurity on EAST with W divertor by an additional radiation losses, the radiative tungsten divertor with Ar impurity seeding for the control of tungsten impurity is studied by SOLPS-ITER simulation. The computational region includes the core, SOL and private flux region. The real geometry and material of plasma facing components (PFCs) are taken into account. The plasma species in the simulation include D 0 , D + 1 , e - , He 0 - He + 2 , C 0 - C + 6 , Mo 0 - Mo + 42 , W 0 - W + 74 and Ar 0 - Ar + 18 . The simulation results show that with the increase of Ar seeding rate at the dome the plasma temperature at the target plates and the concentration of W impurity in the plasma decrease. The simulation plays an important role in the study of radiative tungsten divertor towards the higher heating power and steady state operation.

First of a kind SOLPS-ITER simulations on tokamak a configuration variable (TCV) that include nitrogen have been performed to model recent nitrogen seeded detachment experiments. Based on spectroscopic measurements, a nitrogen recycling coefficient R-p(N) approximate to 0.3-0.5 on the graphite walls of TCV is estimated. The experimentally observed decrease of core nitrogen density with increasing plasma density is reproduced and linked to a reduction of the ionisation mean free path in the scrape-off layer. Although the influence of sputtered carbon impurities from TCV's graphite wall cannot be fully eliminated, seeding nitrogen increases control over the total impurity density. This facilitates disentangling the effect of impurities from that of high upstream density on the main characteristics of detachment, namely target power and ion current reductions and the development of a parallel pressure drop. Increasing the density and the seeding rate reduce the power on the divertor targets in a different way: with density, the ion current increases and the target temperature strongly decreases, whereas seeding impurities decreases the ion current and affects less strongly the temperature. The reduction in ion current when seeding nitrogen is due to a lower ionisation source, which is not related to power limitation nor an increased momentum loss, but to a decrease of the ionisation reaction rate. Impurity seeding leads to less volumetric momentum losses (and hence pressure drop) than density ramps, for the same level of energy flux reduction. Additionally, main chamber sputtering of carbon is identified as a possible explanation for the missing target ion current roll-over during density ramps in the simulations.

© 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

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters lambda(div)(q). Otherwise and further upstream the mid-plane heat decay length lambda(q) parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that lambda(q) upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations lambda(ne)>lambda(Te)>lambda(q) hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of lambda(ne) into account, rather than lambda(q), as the former actually defines the required minimum upstream SOL grid extent. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2018.10.013

Total flux expansion, a divertor magnetic topology design choice embodied in the Super-X divertor, is predicted through simple analytic models and SOLPS calculations to reduce the plasma and impurity density detachment thresholds as the outer divertor separatrix leg position and the strike-point major radius, R-t, are increased. However, those predictions are contradicted by recent TCV experimental results. In this study, utilizing the SOLPS-ITER code, we are able to both match TCV results and demonstrate that the effect of total flux expansion is counteracted by two other divertor geometry design characteristics that affect neutrals: (a) the strike-point angle to the outer target; and (b) the effect of physical baffles that reduce the amount of neutrals escaping from the divertor. We quantify the role of those neutral effects through developing and applying a quantitative definition of neutral trapping. The results of this study indicate that improved divertor design, properly utilizing the three design characteristics discussed should lead all effects to be additive in reducing the detachment threshold. A second implication of this study is that any assessment of alternative topologies must separate out the effects of magnetic topology from neutral design characteristics.

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters lambda(div)(q). Otherwise and further upstream the mid-plane heat decay length lambda(q) parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that lambda(q) upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations lambda(ne)>lambda(Te)>lambda(q) hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of lambda(ne) into account, rather than lambda(q), as the former actually defines the required minimum upstream SOL grid extent.

The possibility to safely operate tokamaks strongly depends on the capability to protect the divertor targets from excessive power load and erosion. Here, we will present a study of the JT-60SA Scenario 3 with a W wall to identify possible criticality in the transition from C to W wall. The assessment of the SOL and divertor plasma conditions is made with the SOLPS-ITER code suite. The simulations results show that safe operations are mainly limited by T e . In addition, the analysis of the balance equations highlights the importance of the impurity injection. It represents the main mechanism to drive the decrease in power load onto the outer target ( P O T ) by impurity radiation and to achieve detachment via the collisions between D + ions and D 2 molecules.