Search

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]

A simplified model describing turbulent plasma transport in the far scrape-off layer has been proposed for calculation of the plasma heat and particle fluxes to the first wall (FW) of the DEMO tokamak. The model has been implemented in the TOKES code for simulation of the heat flux distribution along the tokamak FW. First simulations of the wall heat flux, using this model, predict more pronounced “hot spots” at the top of the DEMO wall in comparison with the exponential heat flux decay model, commonly used for simulation of the flux. [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]

Simulation of the mitigation efficiency of the RE beam impact on the ITER first wall by special sacrificial W diaphragm has been performed using ENDEP, MEMOS and TOKES codes. According to the technique proposed, the RE beam should deposit its energy onto the special sacrificial diaphragm consisting of separate tungsten ‘nails’. The ‘nails’ is partly evaporated and then the beam power deposited into the vaporized and ionized W plasma shield. The wall heat load mitigation is due to the RE beam energy deposition into the W plasma and its re-radiation onto larger wall area. Calculated tolerable beam energy should not exceed 90 MJ to ensure absence of melting for 10 ms RE beam deposition. [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 experiments in the MK-200UG facility dedicated to verify the FOREV-2D simulations of ITER core contamination with carbon vaporized during ELMs has been performed. In these experiments the carbon fibre composite (CFC) of NB31 grade have been treated with plasma heat fluxes relevant for ITER ELMs. The analysis revealed that thin layer of few hundred microns on CFC surface is damaged and its thermoconductivity effectively reduced approximately three times, but the CFC bulk has the reference thermoconductivity. Good agreement between the measured and the calculated profiles for carbon plasma electron density at various hydrogen plasma heat loads as well as the agreement between the measured and the simulated dependences of the absorbed energy density on the applied heat load provide reliable validation of the carbon plasma shields simulated with the FOREV-2D code. High carbon plasma shield densities of 1023–1024 m−3 predicted in the simulations for ELM-produced shields has been proved in these MK-200UG experiments. [Copyright &y& Elsevier]

Abstract: Type I ELMs in ITER would result in the divertor targets vaporization. Vaporized carbon plasma can contaminate the thermonuclear plasma in the ITER core decreasing the plasma temperature and the fusion power or even run the confinement into disruption. Simulations of the tolerable ELM frequency as a function of ELM size are carried out for ITER using TOKES and FOREV-2D codes. Improved FOREV-2D scenario proposed for simulation of the carbon plasma influx into ITER pedestal. An elaborated model for carbon penetration into the confined plasma from pedestal is implemented in TOKES. Simulation of tolerable ELM frequency has been done for 3.5–12.1MJ of the total ELM energy loss. The tolerable frequency is less than 10Hz for the ELM of 7.3MJ and decreases with increasing of the energy loss. [Copyright &y& Elsevier]

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]

Abstract: The interaction of hot plasma with material in fusion devices can result in material erosion and irreversible damage. Carbon based materials are proposed for ITER divertor armour. To simulate carbon erosion under high heat fluxes, electron beam heating in the JUDITH facility has been used. In this paper, carbon erosion under energetic electron impact is modeled by the 3D thermomechanics code ‘PEGASUS-3D’. The code is based on a crack generation induced by thermal stress. The particle emission observed in thermal shock experiments is a result of breaking bonds between grains caused by thermal stress. The comparison of calculations with experimental data from JUDITH shows good agreement for various incident power densities and pulse durations. A realistic mean failure stress has been found. Pre-heating of test specimens results in earlier onset of brittle destruction and enhanced particle loss in agreement with experiments. [Copyright &y& Elsevier]

Exposure of the carbon fibre composites (CFC) NB31 and NS31 by multiple plasma pulses has been performed at the plasma guns MK-200UG and QSPA. Numerical simulation for the same CFCs under ITER type I ELM typical heat load has been carried out using the code PEGASUS-3D. Comparative analysis of the numerical and experimental results allowed understanding the erosion mechanism of CFC based on the simulation results. A modification of CFC structure has been proposed in order to decrease the armour erosion rate. [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]

Abstract: Tungsten plasma transport along the magnetic field calculated with the TOKES code has been validated using dedicated experiment in a plasma gun under the ELM-like conditions. The plasma velocity is in a good agreement with the measured one. The ion composition of the simulated tungsten plasma is in a qualitative agreement with the one reconstructed from the measured spectrum of the tungsten plasma, probably because of rather incomplete database for tungsten ions radiation. Nevertheless, it has been identified the radiation from cold and hot tungsten plasma regions with ion composition maxima at W7+–W8+ and W13+–W16+ correspondingly. [Copyright &y& Elsevier]

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: Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8–10cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J × B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated. [Copyright &y& Elsevier]

Abstract: This paper is concerned with investigation of an erosion of the ITER-like divertor castellated targets of pure tungsten, lanthanum tungsten and CFC under plasma heat loads expected during the Type I ELMs and disruptions in ITER. These experiments were carried out on a plasma gun QSPA-T at the SRC RF TRINITI under EU/RF collaboration. The targets were exposed by series repeated plasma pulses with heat loads in the range of 0.2–2.5MJ/m2 and a pulse duration of 0.5ms. The erosion value as a function of pulse number and energy density were obtained. The erosion of lanthanum tungsten started at the lower energy density as compared with pure tungsten and was mainly due to a melt layer movement and a droplets ejection. Characteristics of ejected droplets were measured. The erosion of CFC macrobrushes under ELM and disruption heat loads was determined mainly by damage of PAN-fibers. [Copyright &y& Elsevier]

Abstract: The behavior of a preheated at 650°C tungsten targets under repetitive ELM-like plasma pulses is studied in simulation experiments with the quasi-stationary plasma accelerator QSPA Kh-50. The targets have been exposed up to 350 pulses of the duration 0.25ms and the surface heat loads either 0.45MJ/m2 or 0.75MJ/m2, which is below and above the melting threshold, respectively. The development of surface morphology of the exposed targets as well as cracking and swelling at the surface is discussed. First comparisons of obtained experimental results with corresponding numerical simulations of the code PEGASUS-3D are presented. [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]

Abstract: This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5–1.5MJ/m2 and pulse duration 0.5ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5μm per pulse. The CFC erosion was negligible at the energy density less than 0.5MJ/m2 and was increased to the average value 0.3μm per pulse at 1.5MJ/m2. The pure tungsten macrobrushes erosion was small in the energy range of 0.5–1.3MJ/m2. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5MJ/m2. [Copyright &y& Elsevier]

Abstract: The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8–10cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated. [Copyright &y& Elsevier]

Abstract: In recent tokamak simulation experiments with the QSPA Kh-50 facility several mechanisms of dust generation from tungsten surfaces under ITER ELM-like energy loads have been identified. Here cracking and melting are reported. The brittle destruction dominates after a few transient impacts when a network of major cracks forms on the surface. Bifurcation of major cracks results in ejection of dust particles with sizes up to ∼30μm. Dust generation occurs also after surface melting and following resolidification when fine crack networks along the grain boundaries develop. In this process the destruction is accompanied by bridge formation due to capillary tension across the fine cracks. Next impacts (even weak melt-free ones) can destroy those bridges, which produces considerable amounts of dust particles of nm-size dust. Surface modification after the repetitive plasma pulses also results in creation of nm-size dust. [Copyright &y& Elsevier]

Abstract: The paper concerns experimental investigations of plasma facing components erosion under the plasma heat loads expected in ITER divertor during transient events such as the Type I Edge-Localized Modes and the disruptions. The experiments were carried out at the TRINITI plasma gun QSPA-T. The carbon fiber composite and tungsten macrobrush targets designed for ITER were exposed to multiple plasma pulses of duration 0.5ms and deposited energy in the range of 0.2–2.5MJ/m2. Between some of the pulses the eroded surface was analyzed with profilometric measurements and electron microscopy. The CFC erosion is determined mainly by damages to the PAN-fibers. While the energy increases from 0.2 to 2.4MJ/m2 the removed layer of PAN-fibers area increases from 0.01 to 10μm per pulse. The erosion of tungsten (pure and lanthanum oxide-doped tungsten) is shown to be determined mainly by crack formation, melt layer movement and droplets ejection. [Copyright &y& Elsevier]