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TOKES simulations of shattered Ne pellets injection into the ITER discharge of 280 MJ thermal plasma energy was carried out taking into account shielding of the pellet. Injection was simulated to be performed through 3 upper port launchers or through 1 equatorial launcher. The shielding of pellets with 2 kPa m3 Ne has minor effects on the radiated power and on wall heat load because the injected amount is much larger than required for radiation. However, the minimum Ne amount, which radiates more than 90% of the core plasma energy (the threshold) is reduced by a factor of 2 compared to the result without shielding. The thresholds are 3 × 16 Pa m3 for upper injection and 62 Pa m3 for the equatorial injector. [ABSTRACT FROM AUTHOR]

Abstract Disruption mitigation by shattered Ne pellet injection (SPI) has been simulated for an H-mode ITER D-T discharge of 280 MJ stored energy using the 3D TOKES code. Heating of the first wall from the radiation flash during the thermal quench of the disruption initiated by the injected pellets has been assessed. The dynamics of Ne plasma expansion in the core with radiation from this plasma have been simulated for the series of pellet sizes from 5.4 × 1024 to 8.6 × 1022 Ne atoms. The threshold pellet size for irradiation of the core energy has been found. The results of SPI and massive gas injection simulations are compared. [ABSTRACT FROM AUTHOR]

Measurements of tungsten (W) plasma shield dynamics have been performed in the 2MK-200 facility under ITER relevant disruptive plasma heat flux densities of ∼100 GW/m 2 . Plasma temperature profiles at two time instants and time dependent electron density profiles are available from these experiments. These data have been used to validate TOKES code simulations of the W plasma shield which is computed to form in front of the ITER W divertor targets during disruptions. The experimental and simulated temperature profiles are in a good agreement for the measurement points at distances of 1–5 cm from the target. An important feature of the W plasma shield seen in the code simulations is the sharp electron temperature drop very close to the target. This region is not resolved in the 2MK-200 facility because of unacceptably low signal/noise ratio. Further experiments in a new, larger plasma gun are planned for experimental validation of this feature. [ABSTRACT FROM AUTHOR]

Simulations of massive gas injection (MGI) into the JET plasma using noble gas (NG) have been performed using the TOKES code. The results of these simulations have been verified by comparison with JET experiment for disruption control. Further simulations performed showed that the amount of NG for triggering the thermal quench can be reduced 40 times comparing with the JET experiment. Such small amount of injected NG should generate runaway electrons (RE) with rather large probability. For mitigation of the wall damage with these RE, a special sacrificial diaphragm consisting of several W ‘nails’ with a characteristic thickness of 2 cm has been proposed. The diaphragm erosion with RE beam has been roughly estimated from above as less than 0.5 cm per disruption. Taking into account the plasma shielding may decrease erosion ∼10 times. [ABSTRACT FROM AUTHOR]

Abstract: Analysis of the available data on massive gas injection (MGI) experiments in JET and comparison with simulation results allows developing the MGI scenario for the TOKES code, which describes the sequence of energy and particle transport processes. Understanding the physics of the processes allows extrapolation of the JET results to ITER. The scenario will be used to simulate the ITER wall heat loads and to estimate the wall damage. [Copyright &y& Elsevier]

Abstract: The integrated tokamak simulation code TOKES is proposed for estimation of the ITER first wall radiation damage during massive injection of noble gas inside the core for mitigation of the disruptions. Simulations of MGI processes ab initio are almost impossible because of complexity of the problem, but one can get reliable results elaborating the TOKES scenario from JET experiments and applying it for ITER conditions. With this aim the TOKES code is used for simulation of JET shots disrupted with MGI. The TOKES scenario for the thermal quench of MGI at JET and its extrapolation to ITER have been developed basing on analysis of the available JET database and on comparison with the simulations performed. [Copyright &y& Elsevier]

Abstract: Brittle destruction of tungsten armour under action of edge localised modes of plasma instabilities (ELMs) in ITER is an important issue determining the lifetime of the divertor. Besides, cracking of the armour produces tungsten dust with characteristic size of 1–10μm flying from the armour surface with velocities up to 10m/s. Influx of the tungsten dust into the ITER confinement decreases the temperature of the plasma, reduces the thermonuclear gain and even may run the confinement into disruption. This paper describes experiments in QSPA-Kh50 plasma gun and modeling, which has been performed for providing more insight into the physics of tungsten cracking under action of ELMs and for confirmation of the important result on stabilization of the crack development at the tungsten armour surface, predicted in our previous paper – the same authors, 2010. The threshold value of energy density deposition for start of tungsten cracking has been measured as 0.3MJ/m2 after 5–10 shots. From analytical considerations three times smaller threshold value has been predicted with increasing number of shots. [Copyright &y& Elsevier]

Abstract: Simulations of tungsten armour cracking under small ELM-like plasma heat load, which does not cause surface melting, have been performed using the PEGASUS-3D code. A dedicated series of experiments have been performed in the QSPA-Kh50 facility for measurements of the unknown tungsten thermophysical properties and for verification of the PEGASUS-3D simulation results. The simulations revealed that a cellular crack network with average mesh size Λ ∼0.5mm formed after first ELMs and the pattern does not change further. With increasing number of repetitive ELMs loads, the average crack width Δ(n) has a maximum value Δ m . The ratio of Δ m /Λ is equal to the tungsten thermal expansion at the maximum surface temperature. Δ(n) tends to this value exponentially. The number of ELMs n m needed for Δ stabilization depends on the ELMs energy density and time duration, n m ∼300 for the simulated ELMs of 0.45MJ/m2 and 0.25ms duration. The PEGASUS-3D code is prepared for simulations of tungsten armour damage under action of ELMs of various energy deposition and time duration. These parameters of ELMs depend on ITER regimes of operation and on how successful will be the efforts on ELMs mitigation. [ABSTRACT FROM AUTHOR]

Abstract: Analysis of the peculiarities of the erosion of NB31 CFC allowed proposition of a new CFC fibre structure for considerable reduction of the erosion rate. The improvement concerns the needling and vowing fibres arrangement only and keeping the same structure of the pitch fibres – the main heat conducting component of NB31. The needling and the vowing fibres of the improved structure provide the armour stiffness, comparable to that of NB31. Numerical simulation of erosion for CFC with the improved structure has confirmed that the erosion rate is four to five times lower in comparison with that of NB31. [Copyright &y& Elsevier]

Carbon fiber composites, CFC, newly developed for the divertor armor of the future tokamak ITER, have a rather complex structure of fiber framework and carbon matrix, allowing CFC to match the thermal conductivity requirements for the tokamak stationary regimes. But for the ITER off-normal events, their behavior is still not investigated in detail. Recent experiments on electron beam facilities simulating the slow transient loads of 20 MW/m2 have shown unusually high erosion rates of CFC in the temperature range of 3000–4500 K. To understand the mechanism of the enhanced erosion, numerical simulations for the new CFC are carried out using the Pegasus code that has been generalized to describe three-dimensional effects of brittle destruction and heat transport in the carbon composites. A new erosion mechanism is discovered. The enhanced erosion is caused by local overheating in CFC materials due to preferential cracking on the fiber surface and thermal isolation of the fibers from matrix. A large difference of fiber and matrix coefficients of thermal expansion is essential for the mechanism. [Copyright &y& Elsevier]

• TOKES numerical simulations of mitigated disruption thermal quenches, caused by SPI in ITER have been performed. • Bremsstrahlung and recombination radiation contributes significantly to the core cooling upon injection of mixed pellet • The minimum Ne amount in two mixed pellets, which radiates >90 % of the core plasma energy has been found. TOKES simulations of mixed Ne/D 2 pellet injection through 1 equatorial launcher into an ITER discharge of 280 MJ thermal plasma energy was carried out. Two kinds of pellets were simulated: 'large', containing 1.1∙1024 D 2 molecules and 'small' with 2.6∙1023 D 2 molecules, both with various amount of Ne and correspondingly lower quantities of D to preserve the pellet size. It was found that the physics of the core energy radiation by the mixed pellet is different from that of the pure Ne pellet. In case of the mixed pellet injection, bremsstrahlung and recombination radiation make a significant contribution to the core cooling along with the Ne line radiation, the latter being the only mechanism for pure Ne pellet injection. The minimum Ne amount in the mixed pellet, which radiates more than 90 % of the core plasma energy (the threshold) has been found to be 1∙1020 Ne atoms for the large pellet and 2∙1021 atoms for the small one. [ABSTRACT FROM AUTHOR]

Abstract Fusion system codes are vital computational tools for the assessment of the key design parameters of a fusion power plant. The main goal is to capture the key physics and technology aspects, pushing the solution to comply with physics and engineering requirements and constraints. Based on that, the magnet system has to be designed to match the requirements on magnetic equilibrium as well as the technological constraints on superconductive coils. A dedicated model was developed to evaluate the coupling effects between the external poloidal field coils (PFCs) and the plasma, described in terms of its magnetic equilibrium configuration. The model first attempts to reproduce a magnetic configuration for a given set of plasma parameters within a target confining region. To achieve such an arrangement, a coil solver calculates the toroidal electrical currents in PFCs. In this paper, the mathematical details of this equilibrium model are presented and the main results referred to magnetic flux, field and coil currents are compared to a reference DEMO scenario calculation. [ABSTRACT FROM AUTHOR]

Highlights: [•] The massive gas injection (neon) is simulated with the two-dimensional tokamak code TOKES assuming the toroidal symmetry. [•] The neon injection, assimilation and transport of impurities through the entire plasma volume are modelled. [•] The output of TOKES is used by the melt motion code MEMOS to assess beryllium wall temperature and the regime with melting. [•] Complete plasma cooling occurs in minimum time of 5.7ms with avoiding Be melting at any point on the first wall. [Copyright &y& Elsevier]

Abstract: Tungsten in the form of macrobrush structure is foreseen as one of the two candidate materials for the ITER divertor and the dome. Melting and thus melt motion and melt splashing are expected to be main mechanisms of metallic target damage that determine the lifetime of ITER plasma facing components. Experiments carried out at the plasma gun facility QSPA-T for ELM-like heat loads demonstrated a significant erosion of frontal and lateral brush edges, which was confirmed by further numerical simulations. In the experiments and numerical simulations a threshold of brush edge melting was determined. In this paper most important mechanisms of melt splashing and melt bridge formation under ITER transient heat loads are analyzed. Approximate criteria for droplet ejection are obtained and the range of transient events without significant droplet injection is calculated. The critical radius of brush edges rounding which prevents the bridge formation at the macrobrush edges is determined. [Copyright &y& Elsevier]