Multidimensional modeling of phenomena and processes occurring during the expansion of the laser-produced plasma for different irradiation conditions related to both the laser beam parameters and the target constructions is a very complex issue, especially when modeling requires consideration of kinetic processes associated with the development of various types of microscopic instability. Multidimensional PIC codes create such a possibility, but their use is limited to modeling phenomena even in a very narrow timescale due to the limited computational capabilities of current supercomputers. For this reason, the paper attempts to interpret the results of the spontaneous magnetic field (SMF) measurements obtained during the PALS (Prague Asterix Laser System) experiment [Pisarczyk et al., AIP Adv. 10, 115201 (2020); Pisarczyk et al., Phys. Plasmas 22, 102706 (2015)] based on the 2D magneto-hydrodynamic (MHD) model [Jach et al., Computer Modeling of Dynamic Interaction of Bodies by Free Point Method (PWN, Warsaw, 2011)]. The MHD equations were used with included arbitrary (i) current of hot electrons treating it as an additional external current and (ii) ion-sound instability responsible for the increase in anomalous resistance in areas with high temperature and low-density plasma. The spatial distribution of magnetic fields and current density obtained from 2D modeling are in acceptable agreement with the experimental results [Pisarczyk et al., Plasma Phys. Controlled Fusion 62, 115020 (2020); Zaraś-Szydłowska et al., AIP Adv. 10, 115201 (2020); Pisarczyk et al., Phys. Plasmas 22, 102706 (2015)]. The inclusion of temporal changes in anomalous resistance in modeling allowed us to explain the persistence of high SMF amplitude at the level of several megagauss after the laser pulse ended due to the effect of magnetic field freezing.

The current development state of the finite difference, multi-fluid, 3D plasma code Findif is detailed. The code was run on four meshes prepared for the OP 1.1 wall geometry of the Wendelstein 7 X stellarator. The meshes were produced for 4 magnetic configurations; two of them are finite-beta (non vacuum). The simulated volume covers plasma edge; the computations of limiter heat load distributions were the main goal. Plasma radiation was not taken into account.

In this work the scenario #2 for the JT-60SA with a carbon divertor configuration is analysed by using the core-edge integrated COREDIV code. A combined seeding with two impurities, high and low atomic number, is proposed for controlling the core and the scrape-off layer radiation and limiting the total power delivered to the divertor plate. As a proof-of-principle tungsten and argon seeding is investigated due to successful results reported for the same scenario with W divertor. Replacing non-recycling W by recycling Xe shows the influence of the recycling coefficient on the final solution. Also the location of the puffing nozzle was found to be a crucial factor for the final result.

A tool for automated mesh production for 3D multifluid plasma transport code Findif is presented. Mesh points for the code lie on magnetic field lines, which, in general, form a complicated tangle. Open field lines that end on solid parts of the machine are the source of difficulties. These lines are usually short and thousands of them are needed. A tool that helps to pick lines for the mesh that is described in this paper is based on calculation the distances of lines already admitted to the mesh and candidate ones. The results of the code run for the limiter configuration (OP-1.1 experimental phase) of the Wendelstein-7X device are shown here. Reasonably even coverage of space by points is achieved.

JT-60SA reference design scenarios at high (# 3) and low (# 2) density have been analyzed with the help of the self-consistent COREDIV code. Simulations results for a standard C wall and full W wall have been compared in terms of the influence of impurities, both intrinsic (C, W) and seeded (N, Ar, Ne, Kr), on the radiation losses and plasma parameters. For scenario # 3 in a C environment, the regime of detachment on divertor plates can be achieved with N or Ne seeding, whereas for the low density and high power scenario (# 2), the C and seeding impurity radiation does not effectively reduce power to the targets. In this case, only an increase of either average density or edge density together with Kr seeding might help to develop conditions with strong radiation losses and semi-detached conditions in the divertor. The calculations show that, in the case of a W divertor, the power load to the plate is mitigated by seeding and the central plasma dilution is smaller compared to the C divertor. For the high density case (# 3) with Ne seeding, operation in full detachment mode is predicted. Ar seems to be an optimal choice for the low-density high-power scenario # 2, showing a wide operating window, whereas Ne leads to high plasma dilution at high seeding levels albeit not achieving semi-detached conditions in the divertor.

The paper is aimed at optimization of parameters of a copper plasma jet produced at the DPF-1000U device, in which the inner electrode face was conically shaped. Preliminary information was obtained by numerical simulations of the plasma jet creation for different copper cones with the use of the two-dimensional magneto-hydrodynamic code KAROL. The simulations suggested that the cone height in the range of 4-7 cm should ensure a good plasma jet quality. The experimental data delivered by means of a 16-frame laser interferometer and a four-frame X-ray pinhole camera fully confirmed this conclusion. In the paper, we demonstrate the results for a 5 cm height cone. The eroded Cu plasma, swept up by the deuterium plasma sheath, was accelerated axially and compressed to very small diameter (3 mm) with an electron density of 7 x 10(18) cm(-3). The Cu plasma jet achieved a velocity of 5 x 10(7) cm/s and reached in the period of about 230 ns a distance (length) of 7 cm. The above results prove a successful adaptation of the plasma focus device to the metallic plasma jet generator.

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.

An analysis of radiative power exhaust for the JT-60SA tokamak with a tungsten divertor is performed with the help of the self-consistent, core-edge integrated COREDIV code. Two scenarios of operation (low and high density) were investigated in the scope of different parameters (electron density at the separatrix and the perpendicular transport in the scrape-off layer) with impurity seeding (Ne and Kr). The calculations show that in the case of the tungsten divertor the power load to the divertor plate is mitigated and the central plasma dilution is smaller compared to the carbon divertor. In the most cases the energy flux through the separatrix is above the L–H transition threshold. For the high density case with neon seeding operation in full detachment mode is observed. Changing the diffusion coefficient in the SOL has a strong influence on the result of the calculations as increased radial transport causes stronger screening effect. Also by changing the electron density on the separatrix the influx of heavy impurities (W, Kr) into the core region can be reduced. The results demonstrate that it is easier to achieve sustainable conditions in the divertor region for the high density scenario, whereas for the low density one reducing the auxiliary heating power seems unavoidable to prevent damaging of the target plate, even for strong seeding gas influx.

JT-60SA design scenarios have been analyzed with the help of the self-consistent core-edge COREDIV code, with the aim to assess the influence of impurities on the plasma parameters and tokamak performance. In particular, the reduction of divertor target power load due to radiation of sputtered and externally seeded impurities has been investigated. For all scenarios considered, the gradual replacement of carbon by low Z seeding impurity (N, Ne) is observed as the gas influx increases. For high auxiliary power and low density scenarios, the carbon and seeding impurity radiation does not effectively reduce power to plate. Consequently, results with very high Zeff ?> (about 6–8) and impurity concentrations (>9%) are observed. For these scenarios, seeding of the high Z impurity like krypton seems to be more appropriate, in particular when the plasma density is increased.

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2019 and is expected to give substantial contributions to both ITER and DEMO scenario optimization. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving the elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the main activities undertaken are presented and summarized.

JT-60SA reference design scenarios at high (#3) and low (#2) density have been analyzed with the help of the self-consistent COREDIV code. Simulations results for standard carbon wall and the full W have been compared in terms of the influence of impurities, both intrinsic (C, W) and seeded (N, Ar, Ne, Kr) on the radiation losses and plasma parameters. For the scenario #3 in carbon environment the regime of detachment on divertor plates can be achieved with N or Ne seeding, whereas for the low density and high power scenario (#2), the carbon and seeding impurity radiation does not effectively reduce power to the targets. In this case only increase of neither average density or edge density together with Kr seeding might help to develop conditions with strong radiation losses and semi-detached conditions in the divertor. The calculations show that in case of tungsten divertor the power load to the plate is mitigated by seeding and the central plasma dilution is smaller compared to the carbon divertor. For the high density case with neon seeding operation in full detachment mode is observed. Ar seems to be an optimal choice for the low density high power scenario #2, showing wide operating window, whereas Ne leads to high plasma dilution at high seeding levels albeit not achieving semidetached conditions in the divertor.

In this paper numerical simulations with the self-consistent COREDIV code of the planned JET DT experiments have been performed. First, record shot from the 1997 experiments was simulated and good agreement with experimental data has been found. Direct extrapolation of the carbon wall results to the new ILW configuration (discharge parameters as for the shot #42746) shows very good core plasma performance with even higher fusion power but with too large power to the divertor. However, with the neon seeding the heat load and plate temperatures can be efficiently reduced keeping good the plasma performance. Investigations have been done also for the planned DT operation scenario based on a conventional ELMy H-mode at high plasma current and magnetic field. Simulations for the reference ELMy H-mode shot #87412 show good agreement with the experimental data but the direct extrapolation of the DD results to deuterium-tritium operation shows relatively poor performance in terms of the achieved fusion power. The situation improves, if the highest heating power is assumed (41 MW) and fusion powers in the excess of 12 MW can be achieved. All the high performance shots require the heat load control by neon seeding which shows rather beneficial effect on the plasma performance allowing for relatively wide operational window in terms of the amount of the allowed neon influx.

Scenario modelling for the demonstration fusion reactor (DEMO) has been carried out using a variety of simulation codes. Two DEMO concepts have been analysed: a pulsed tokamak, characterized by rather conventional physics and technology assumptions (DEMO1) and a steady-state tokamak, with moderately advanced physics and technology assumptions (DEMO2). Sensitivity to impurity concentrations, radiation, and heat transport models has been investigated. For DEMO2, the impact of current driven non-inductively by neutral beams has been studied by full Monte Carlo simulations of the fast ion distribution. The results obtained are a part of a more extensive research and development (R&D) effort carried out in the EU in order to develop a viable option for a DEMO reactor, to be adopted after ITER for fusion energy research.

This paper aims at investigation of efficiency of an ablative plasma energy transfer into a massive aluminum target using different atomic number ablators. For this reason, several target materials representing a wide range of atomic numbers (Z = 3.5-73) were used. The experiment was carried out at the iodine Prague Asterix Laser System. The laser provided a 250 ps pulse with energy of 130 J at the third harmonic frequency (lambda(3) = 0.438 mu m). To study the plasma stream configurations a four-frame X-ray pinhole camera was used. The electron temperature of the plasma in the near-surface target region was measured by means of an X-ray spectroscopy. The efficiency of the plasma energy transport to the target was determined via the crater volume measurement using the crater replica technique. The experimental results were compared with two-dimensional numerical simulations where the plasma dynamics was based on the one-fluid, two temperature model, including radiation transport in diffusive approximation and ionization kinetics. It was shown that the plasma expansion geometry plays an important role in the ablative plasma energy transfer into the target.

The several Demo design concepts have been analyzed using the code Corediv solving self consistently the core and scrape off layer regions. The numerical analyses show that there exists the limit on the concentration of impurities above which the solution with burning plasma does not exists. This limits reduce the possibility of reduction of power loads to target plate. In the paper the admissible range of seeded impurity concentration introduced to reduce the power load to divertor plate is analyzed. The possible mechanisms leading to the limitation are shown. (© 2014 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

The effect of magnetic field line ergodization that eliminates magnetic surfaces (either by a resonant magnetic perturbation like in TEXTOR-DED or by intrinsic plasma effects like in W7-X) imposes the need for plasma transport models being able to describe this properly. To handle the ergodicity the concept of local magnetic coordinates allowing a correct discretization of the transport equations with minimized numerical errors is used. For the simulation of plasma transport in perturbed volume, a numerical method based on the finite difference concept has been developed, using a custom-tailored unstructured grid in local magnetic coordinates. This grid is generated by field line tracing to guarantee complete separation of the large parallel transport along B and that perpendicular to B and the ergodicity of the magnetic field does not limit applicability of the method in contrast to the methods based on finite volume ansatz. Perpendicular and parallel fluxes can be effectively separated in our approach and treated independently in the numerical method which has been implemented in the FINDIF code.The finite difference code FINDIF is used to investigate the energy transport in the complex 3D TEXTOR-DED tokamak geometry, where the plasma structures and transport are closely related to the structure of the magnetic field lines. Numerical grids have been prepared in order to simulate 12/4 and 6/2 modes of the DED operation, respectively. In particular, the question, what is the role of long and short magnetic field lines in the heat transfer from the core plasma to divertor surface, is addressed. Simulation results are compared with experimentally determined temperature profiles and heat fluxes at the target.

Results are presented from two-dimensional MHD simulations of X-pinch implosion. The simulations were performed in the (r, z) and (x, y) geometries for homogeneous (dense plasma) and heterogeneous (core-corona) loads. The formation of a minidiode, the development of a neck and an X-radiating hot spot, and the influence of the plasma corona on the implosion dynamics of the dense X-pinch plasma were investigated. For through simulations, the conical neck model was used, whereas a detailed analysis of the X-ray burst was performed in the parabolic neck model. The MHD processes occurring during the implosion of oblique shock waves and the onset of instability of the plasma column were examined. It is found that, due to the quasi-periodic character of these processes, the neck compression proceeds in a cascade fashion. The plasma state in a hot spot just before the break of the neck is analyzed, and the possibility of generating fast particle beams is considered.

Plasma-Focus (PF) devices, which are based on high-voltage high-current pulse discharges, belong to the noncylindrical Z-pinches. They can produce high-temperature dense magnetized plasma and intense radiation pulses (of Xrays, electrons, ion beams and fusion products). The paper reports on numerical modeling of different phases of high-current pulse discharges in a PF-type device. The numerical results concerning the breakdown and collapse phases are presented and discussed.

This paper presents results of the recent plasma-focus (PF) experiments carried out with PF facilities, which was operated at energies ranging from 0.5 MJ to about 1 MJ. Particular attention has been paid to pinch evolution, the emission of pulsed X-ray, fast electron beams, and fusion produced neutrons. Some theoretical models of the initial breakdown, which occurs at the insulator surface, are compared. It is pointed out that modeling of the breakdown is sensitive to kinetics of ionization processes and transport coefficients. Progress in experimental studies of the axial acceleration phase is unsatisfactory. Important experimental data have been collected, but new measurements are still needed. For the radial collapse phase, it was shown that the MHD modeling is efficient until the maximum compression, but plasma instabilities require more sophisticated approaches. The pinch phase was investigated by means of different diagnostics. Fusion neutron yields were measured in different experiments, but some discrepancies in scaling must still be explained. The conclusions concern directions for further studies and optimization of large-scale high-current PF facilities.

The numerical data on electrical explosion of X-pinches have been obtained to analyze the physical phenomena, which lead to a generation of short high-intensity X-ray pulses observed in experiments. The main attention is given to MHD cumulative processes occurring in point-like regions of a non-equilibrium compressed plasma with high temperatures. The model gives the scaling laws for X-pinch cascade evolution, in particular, for hot-spot plasma parameters and X-ray burst. Namely, the duration of an x-ray pulse is of several picosecond, the size of a spark region is not larger than a few microns, and the X-ray pulse intensity is about 10(17)W/cm(2).