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X-Point Radiator (XPR) regimes have been obtained in WEST tokamak experiments with nitrogen seeding during the experimental campaigns of 2023 and 2024. These experiments showed the formation of a stable toroidal radiating ring near the X-point, similar to observations in other devices such as JET, ASDEX-Upgrade, TCV, and COMPASS. In WEST, the onset of this regime is associated with a sharp transition of the divertor plasma from hot to cold and dense conditions, with increased particle fluxes, indicating that the plasma is not detached. At the same time, core conditions are significantly improved. These scenarios were successfully controlled in WEST using an interferometry line-of-sight passing through the X-point. Interpretative modeling of these discharges with the SOLEDGE3X-EIRENE code reveals that a physics mechanism needed to stabilize WEST nitrogen XPRs is not present when driving the simulations at constant power. On the contrary, a stable XPR can be obtained by increasing the power injected at the time of the XPR onset to represent the reduction of W contamination, highlighting the need to describe the plasma dynamics and go toward integrated core–edge simulations.

The effect of thermodynamic nonequilibrium of hydrogen isotope recycling, which consists in significant deviation of the velocity distribution function (VDF) of neutral hydrogen isotope atoms from the Maxwellian VDF, on the passive signal of the charge-exchange recombination spectroscopy (CXRS diagnostics) of edge plasma in tokamaks is analyzed using the ITER tokamak reactor as an example. The calculations of the VDF for atomic and molecular deuterium are carried out using the three-dimensional Monte Carlo code EIRENE as a part of the two-dimensional transport code SOLPS and the semi-analytical Ballistic Model BM1D2V, which is one-dimensional in coordinate and two-dimensional in projections of atomic velocity. The source of level population of the Be IV beryllium ion, the line radiation of which is used in the CXRS diagnostics, is calculated for the above two methods of calculating the VDF and the average atomic kinetic energy, as well as in the case of neglecting the thermodynamic nonequilibrium of recycling, when the temperature of atoms is equal to the temperature of the plasma ions. The calculations of the cross section for the charge-exchange reaction of beryllium ions Be V on deuterium atoms using the ARSENY code and other codes and the erosion kinetics of the beryllium first wall using the three-dimensional code ERO2.0 are used. The calculations of the level populations of the Be IV beryllium ion using the nl-KinRyd code, the luminosity of the ions, and the passive charge-exchange radiation intensity on the observation chord are carried out for the three indicated cases of the VDF of deuterium atoms. A significant dependence of the results of the conducted predictive modeling on the thermodynamic nonequilibrium of recycling is shown. [ABSTRACT FROM AUTHOR]

The dithering H-mode phase, characterized by oscillations, is generally observed at input power values close to the L-H transition power threshold and low plasma collisionalities (low electron density and/or high plasma temperature). Measurements to characterize the dithering phase are presented for the low aspect ratio, high magnetic field tokamak, ST40. The dithering phase oscillation frequency is observed between 400 and 800 Hz and demonstrates an inverse relationship with core plasma density. Dithering phase H-modes are documented across a nonlinear, low-density power threshold operational space, with signature low- and high-density branches. The minimum power threshold for dithering H-mode access is measured at a core, line average electron density of 4.7(±0.5)×1019 m−3, close to a predicted value of 4.1(±0.4)×1019 m−3 from multimachine studies. ASTRA calculated values of power coupled to the ion species, at the dithering H-mode transition, exhibit a similar nonlinear dependence on density. This analysis points to the important contribution of the ion thermal channel to the L-H phase transition. The low-frequency plasma density and D-alpha dithers appear to be accompanied by sudden bursts of magnetohydrodynamic (MHD) activity. A simple model is tested to demonstrate a possible scenario of self-regulation among turbulence, zonal flows, pressure (density) gradient and MHD activities. [ABSTRACT FROM AUTHOR]

We report on comprehensive experimental studies of turbulence spreading in edge plasmas. These studies demonstrate the relation of turbulence spreading and entrainment to intermittent convective density fluctuation events or bursts (i.e. blobs and holes). The non-diffusive character of turbulence spreading is thus elucidated. The turbulence spreading velocity (or mean jet velocity) manifests a linear correlation with the skewness of density fluctuations, and increases with the auto-correlation time of density fluctuations. Turbulence spreading by positive density fluctuations is outward, while spreading by negative density fluctuations is inward. The degree of symmetry breaking between outward propagating blobs and inward propagating holes increases with the amplitude of density fluctuations. Thus, blob-hole asymmetry emerges as crucial to turbulence spreading. These results highlight the important role of intermittent convective events in conveying the spreading of turbulence, and constitute a fundamental challenge to existing diffusive models of spreading.

The mathematical model of a rarefied edge plasma is considered in this paper. The Vlasov–Poisson equations are solved based on the developed algorithm of successive iterations in time after a pulse change in one or more parameters of the problem. The problem is multidimensional, nonstationary, and multiparameter. The developed fundamental special techniques make it possible to carry out calculations without the use of supercomputers. Mathematical models of the edge plasma parameters near a flat structural element of a spacecraft moving in the ionosphere along a circular orbit are reported. [ABSTRACT FROM AUTHOR]

Estimates of plasma conditions in the far scrape-off layer (SOL) and of first wall (FW) fluxes in ITER are key input parameters to first wall erosion and impurity migration models, which are in turn involved in the assessment of FW panel lifetime and fuel retention studies. SOLEDGE3X up-to-the wall boundary plasma simulations are performed for ITER, based on an expected Pre-Fusion-Power-Operation (PFPO-1) scenario at PSOL= 20MW, including an impact study of enhanced far-SOL transport. This latter study concerns the possible formation of density shoulders, which are modelled here by applying an increase to the prescribed perpendicular particle and heat diffusivity coefficients maps in the far-SOL in the code, in order to flatten the density and temperature profiles there. Several kinds of such obtained “shoulders” are considered. A brief comparison with SOLPS-ITER is performed on the reference case with uniform coefficients, and shows good agreement. When far-SOL transport is increased, temperatures computed on the first wall rise to 20–30 eV for ions, and to 10–20 eV for electrons. It is also found that for first wall quantities of interest in the ITER machine, the assumed level of perpendicular transport in the far-SOL is the most relevant parameter, with the location at which transport is increased being much less important.

ITER Pre‐Fusion Power Operation 1 (PFPO‐1) phase low‐power plasmas are simulated with the SOLEDGE3X‐EIRENE code, a multi‐fluid edge plasma solver that can describe plasma conditions up to the machine's first wall, including a kinetic description of neutrals. Here, SOLEDGE3X is used in 2D mean‐field mode. A throughput scan is performed, reproducing a similar scan from the ITER SOLPS simulations database. The physics assumptions for these simulations are presented, in particular the plasma–neutral interaction model where molecule charge‐exchange and ion‐molecule elastic collisions have newly been added. Challenges encountered in such up‐to‐the‐wall simulations are discussed, along with remaining open questions. Results are presented with two foci: on targets, and on the rest of the first wall. On targets, conditions transition from attached to partially detached as expected, and toroidally symmetric heat fluxes (i.e. without 3D effects) remain below 5 MW/m2, within the divertor design limits. An increase in the parallel heat flux decay length λq with throughput is also observed. On the rest of the first wall, increasing puff rate does not only spread power deposited on the target but also increases heat load on the first wall. These fluxes remain however small (<100 kW/m2). [ABSTRACT FROM AUTHOR]

The multi-component fluid closure derived by Zhdanov (2002 Transport Processes in Multicomponent Plasma (London: Taylor and Francis)) is implemented in the fluid code SOLEDGE3X-EIRENE to deal with arbitrary edge plasma composition. The closure assumes no distinction between species such as light versus heavy species separation. The work of Zhdanov is rewritten in a matricial form in order to clearly link friction forces and heat fluxes to the different species velocities and temperature gradients. [ABSTRACT FROM AUTHOR]

The impact of generalized sheath boundary conditions, taking into account the sheath perturbations driven by the Alfvén waves, on the propagation of plasma filaments at the tokamak edge is considered. A model describing the filament dynamics and containing new boundary conditions is developed. Estimates of the model control parameter—the sheath impedance—on the plasma transport are obtained. Simulations of the filament dynamics under the conditions characteristic of edge plasmas in NSTX and Alcator C‑Mod tokamaks are performed. It is shown that the sheath impedance leads to an increase in the mean values of the electrostatic potential and advection velocities of filaments, in agreement with experimental observations, requiring, at the same time, a more detailed description of the Alfvén wave reflection from the conducting surfaces of the installations. [ABSTRACT FROM AUTHOR]

The plasma-facing material is one of the most critical issues in fusion devices, which is important for the improvement of the life time of divertor targets, the control of impurities and the main plasma parameters. The high- energy plasma results in the high particle flux and heat flux deposition on the divertor target, leading to the erosion o f target plates. The re fore, the materials of target plates have an effect on the edge plasma parameters. In this work, the distributions of electron density, electron temperature, and neutral particle density with different divertor materials ( lithium, boron, carbon, and tungsten) are investigated with the help o f the three-dimensional edge transport program EMC3-EIRENE. It is found that the downstream electron density with tungsten targets is significantly lower than that of the other three low-Z target mate ria ls, and the downstream electron temperature of the tungsten targets is higher. Similar distributions of the electron density and temperature are also observed in the mid- p lane region. [ABSTRACT FROM AUTHOR]

Background The energy produced by nuclear fusion on a Tokamak device is mainly exhausted through the divertor, its service life is directly affected by the interaction between huge heat flux from core and the divertor target. The large amount of impurity produced by the heat flux hitting the target leads to the reduction of the plasma confinement performance whilst pumping is an important means to control plasma density and impurity density. Purpose This study aims to investigate the influence of pumping on the heat load of the target plate which is of reference significance for the future experiment. Methods Based on the experimental parameters of the HL-2A Tokamak, SOLPS-ITER code was used to study the effect of pumping on the heat load of the divertor target under different upstream electron densities. Analysis was performed through density scanning to find the sensitive threshold whilst and atom-molecular collision process was applied to the effect of pumping on the distribution of plasma and neutral particle parameters in divertor region at different upstream electron density. Results Density scanning results show that pumping near the detachment threshold ( T e t O S P ~5 eV) has a greater effect on the thermal load of the target plate. When the pumping rate is 12 m3·s-1, 36 m3·s-1 and 96 m3·s-1 respectively, the miss threshold and thermal load peak of outer target plate are 1.11, 1.24, 1.39 and 1.37, 1.96, 2.54 times of those without pumping respectively. Conclusion It is found that the decreases of deuterium molecular density results in the energy of the collision reaction power decreases when the upstream electron density exceeds the detachment threshold, leads to the increase of the temperature and energy flow of the plasma in the target plate.

The impact of triangularity on edge plasma transport and turbulence is addressed from full 3D turbulence simulations performed with TOKAM3X. Flux driven fluid simulations are run on analytical magnetic equilibria generated with positive and negative triangularity δ in a bottom limiter configuration. The conservation of the energy is assured by the increase of the bottom limiter radial position from δ>0to δ

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.

In this work, evolution of dust grains from different materials used in fusion energy installations was studied and a model was constructed that describes the generation of dust. The model accounts for the thermochemical, electrical, and other properties of the materials of the fusion reactor wall. It was shown that the dominant process that leads to the decrease in dust grain mass is the thermal evaporation, which is determined by the saturated vapor pressure at thermal equilibrium temperature. Estimates of the lifetime of dust grains from different materials were obtained depending on plasma parameters. The findings of this work can be useful for estimates of the length of penetration of dust grains into the reactor. The difference in the dynamics of particles from light and heavy elements was shown. Out of the four elements under study (Be, Ni, Mo, and W), nickel grains demonstrate the highest penetrating ability due to their long lifetime and moderate weight. [ABSTRACT FROM AUTHOR]

A new type of scrape-off layer plasma instability associated with radial inhomogeneity of plasma density, contact of plasma with conducting material surfaces, and neutral recycling is presented. In order to elucidate the physics of this instability, the simplest possible setup is analyzed. [ABSTRACT FROM AUTHOR]

This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between seeded (Ar) and intrinsic impurities (tungsten, helium). This work is to analyse the influence of the helium and hydrogen recycling on the He confinement time in DEMO reactor. Simulations have been done for argon seeding puff 5.0 × 10211/s. It is found that recycling coefficient of helium have strong influence on the He confinement time, which increases from 7.25 s to 26.9 s going from lowest to highest recycling coefficient, but it has small influence on the effective charge state and radiation in the SOL. Alpha power decreases only by about 20%, which is the effect of main plasma dilution. [ABSTRACT FROM AUTHOR]

Highlights • The COREDIV code has been used to simulate DEMO inductive discharges with different impurity seeding (Ar, Xe) and mix gasses. • For xenon seeding for case with standard radial diffusion in SOL, such regime of operation seems not to be possible. • For Ar and combined Ar+Xe seeding it is possible to achieve H-mode plasma operation with acceptable level of the power to the target, but is upper limit on Xe concentration. • The influence of the prompt re-deposition model on the DEMO working point is relatively weak. Abstract In this paper numerical simulations with COREDIV code, which self-consistently solves radial transport equations in the core region and 2D multi-fluid transport in the SOL of EU DEMO discharges in full tungsten environment (W divertor and wall) for H-mode scenario has been performed with different impurity seeding. The optimal impurity mix for reactor would require the use of high Z impurity (low dilution, higher P α) in core and low/medium Z impurity radiating in the SOL region. In this paper we focus on investigations how the operational domain of EU DEMO can be influenced by different seeding (Ar, Xe) and mix gasses. In case of Xenon seeding, H-mode operation can NOT be achieved for standard radial diffusion (D SOL = 0.5m2/s) in the SOL. In the simulation, we have considered combined seeding of Ar (good radiator in SOL) and Xe (good radiator in core). In order to find the optimal impurity mix, simulations have been performed in such a way that for a few fixed levels of Xe puff, Ar puff has been increased from zero to the maximum value allowed by the code stability. For combined seeding: xenon + argon working point for EU DEMO can be found. [ABSTRACT FROM AUTHOR]

Abstract The long-term operations of the EAST experiments result in a substantial redeposition of impurities on the first wall, which is considered as the potential impurity source during the discharge. The transport properties of lithium impurity in the scrape-off layer (SOL) of EAST have been studied by the three-dimensional (3D) edge transport code, the EMC3-EIRENE. The neutral lithium impurities from four impurity source positions on the first wall have been used in the modelling. The detailed analysis of the relationship between the impurity source positions and the density profiles of the lithium impurity for different charge states has been carried out. It is found that the density profiles of the Li1+ ions are determined by the impurity source positions, whereas the density profiles of the Li2+ and Li3+ ions are not related to the first wall source positions. [ABSTRACT FROM AUTHOR]

Numerical scan at constant β shows that core and scrape‐off layer (SOL) radiations do not depend on the plasma current (Ip). Whereas the SOL radiation increases with seeding, the core radiation, however, does not continue to increase with seeding but rolls over at higher seeding rates in the simulations. The core plasma contamination by W ions is low, cW ≪ 10−4. When the seeding starts, an increase in radiation power leading to a reduction in Ploss = (Paux − Prad) is observed, influencing the plasma confinement. The power scan at constant Ip indicates that the core radiation, Pplate, PSOL (and even SOL radiation), saturates with seeding. In addition, strong dilution with increasing seeding (Zeff ≫ 3) and large W concentrations with increasing power are found. Comparing neon with nitrogen seeding, it is seen that neon leads to slightly larger total radiation than nitrogen. However, that is achieved with much higher plasma contamination (Zeff ∼ 4–5) and dilution in the case of Ne, and simultaneously the power crossing the separatrix is lower for Ne than for N, indicating better H‐mode performance in N2‐seeded discharges. [ABSTRACT FROM AUTHOR]

Edge plasma properties have been modified in the Saskatchewan Torus-Modified tokamak by means of resonant magnetic perturbations (RMP). It has been found that the radial profiles of ion saturation current and floating potential in the edge region can be modified by an externally applied static (m = 2, n = 1) RMP field. An increase in the pedestal plasma density (n) and more negative electric field (Er) have been observed in the plasma edge region. It is believed that the RMP field altered the plasma transport in the edge and scrape-off layer regions, leading to a higher density pedestal and a potential drop in some cases. During the enhanced confinement phase, it is possible to identify a region where intermittent transport events, the so-called blobs, are created and the holes of lower density left behind. [ABSTRACT FROM AUTHOR]

In this work, we use a method based on the concept of particle confinement time uniqueness to calculate the electron density and temperature in ohmically heated, edge plasma of the Hefei tokamak-7. Here, with the help of the data taken from Johnson and Hinnov's table, we have done an extensive work to find electron densities and temperatures that satisfy the uniqueness to evaluate the temporal evolution of electron density and temperature . The results are in good agreement as measured from the Langmuir probe array in previous works. [ABSTRACT FROM AUTHOR]

A two dimension-in-space and three dimension-in-velocity (2d3v) Particle-In-Cell (PIC) code is applied to investigate the plasma behaviors at the divertor gaps region in this work. Electron and D + ion fluxes to the tile surface in the poloidal and toroidal gaps for different shaped edges are compared to demonstrate the optimized tile geometry. For poloidal gap, shaped edge in the shadowing side makes more ions penetrate into the gap, while shaped edge in the wetted side can mitigate the peak flux value. For toroidal gap, most ions entering the gap impinge on the side tile mainly due to the E × B drift, and shaped wetted edges also can mitigate the peak heat fluxes. In addition, effects of magnetic field inclination angle from toroidal direction on the plasma behaviors are simulated for poloidal and toroidal gaps, respectively. It is found that the magnetic field angles don’t influence the plasma behaviors in poloidal gap; while significant changes have been observed in the toroidal gap. [ABSTRACT FROM AUTHOR]

ITER expects to produce fusion power of about 0.5GW when operating with tungsten (W) divertor and beryllium (Be) wall. The influx of W from divertor can have significant influence on the discharge performance. This work describes predictive integrated numerical modeling of ITER discharges using the COREDIV code, which self-consistently solves the 1D radial energy and particle transport in the core region and 2D multi-fluid transport in the SOL. Calculations are performed for inductive ITER scenarios with intrinsic (W, Be and He) impurities and with seeded impurities (Ne and Ar) for different particle and heat transport in the core and different radial transport in the SOL. Simulations show, that only for sufficiently high radial diffusion (both in the core and in the SOL regions), it is possible to achieve H-mode mode plasma operation (power to SOL > L-H threshold power) with acceptable low level of power reaching the divertor plates. For argon seeding, the operational window is much smaller than for neon case due to enhanced core radiation (in comparison to Ne). Particle transport in the core characterized by the ratio of particle diffusion to thermal conductivity) has strong influence on the predicted ITER performance. [ABSTRACT FROM AUTHOR]

Numerical simulation with COREDIV code of DEMO H-mode discharges (tungsten divertor and wall) are performed considering the influence of seeding impurities with different atomic numbers: Ne, Ar and Kr on the DEMO scenarios. The approach is based on integrated numerical modeling using the COREDIV code, which self-consistently solves radial transport equations in the core region and 2D multi-fluid transport in the SOL. In this paper we focus on investigations how the operational domain of DEMO can be influenced by seeding gasses. Simulations with the updated prompt re-deposition model implemented in the code show that only for Ar and Kr, for high enough radial diffusion in the SOL, it is possible to achieve H-mode plasma operation (power to the SOL> L-H transition threshold power) with acceptable level of the power to the target plates. For neon seeding such regime of operation seems not to be possible. [ABSTRACT FROM AUTHOR]

Due to the extreme conditions required to produce energy by nuclear fusion in tokamaks, simulating the plasma behavior is an important but challenging task. We focus on the edge part of the plasma, where fluid approaches are probably the best suited, and our approach relies on the Braginskii ion–electron model. Assuming that the electric field is electrostatic, this yields a set of 10 strongly coupled and non-linear conservation equations that exhibit multiscale and anisotropy features. The computational domain is a torus of complex geometrical section, that corresponds to the divertor configuration, i.e. with an “X-point” in the magnetic surfaces. To capture the complex physics that is involved, high order methods are used: The time-discretization is based on a Strang splitting, that combines implicit and explicit high order Runge–Kutta schemes, and the space discretization makes use of the spectral element method in the poloidal plane together with Fourier expansions in the toroidal direction. The paper thoroughly describes the algorithms that have been developed, provides some numerical validations of the key algorithms and exhibits the results of preliminary numerical experiments. In particular, we point out that the highest frequency of the system is intermediate between the ion and electron cyclotron frequencies. [ABSTRACT FROM AUTHOR]

The new code TOKAM3X simulates plasma turbulence in full torus geometry including the open field lines of the Scrape-off Layer (SOL) and the edge closed field lines region in the vicinity of the separatrix. Based on drift-reduced Braginskii equations, TOKAM3X is able to simulate both limited and diverted plasmas. Turbulence is flux driven by incoming particles from the core plasma and no scale separation between the equilibrium and the fluctuations is assumed so that interactions between large scale flows and turbulence are consistently treated. Based on a domain decomposition, specific numerical schemes are proposed using conservative finite-differences associated to a semi-implicit time advancement. The process computation is multi-threaded and based on MPI and OpenMP libraries. In this paper, fluid model equations are presented together with the proposed numerical methods. The code is verified using the manufactured solution technique and validated through documented simple experiments. Finally, first simulations of edge plasma turbulence in X-point geometry are also introduced in a JET geometry. [ABSTRACT FROM AUTHOR]

The modeling of the ITER edge is performed with the use of the code B2SOLPS5.2 in the presence of the electron conductivity caused by RMPs as well as for the reference case with the same input parameters but without RMPs. The radial electric field close to the neoclassical one is obtained without RMPs. Even the modest level of RMPs changes the direction of the electric field and causes the toroidal spin-up of the edge plasma. At the same time the pump-out effect is small. (© 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim) [ABSTRACT FROM AUTHOR]

Power exhaust is one of the major challenges that future devices such as ITER and DEMO will face. Because of the lack of identified scaling parameters, predictions for divertor plasma conditions in these devices have to rely on detailed modelling. Most plasma edge simulations carried out so far rely on transport codes, which most of the times consist of a fluid code for the plasma coupled to a kinetic Monte Carlo (MC) code for neutral particles. One of the main difficulties in interpreting code results is the statistical noise from the MC procedure, which makes it difficult to define a convergence criterion for the simulations. In this work, we elaborate on similarities between noisy transport code simulations and turbulence simulations, and argue that the time averaged solution is a well defined stationary solution for the system. We illustrate these ideas with a simple slab test case with fluid neutrals, to which we add synthetic noise. In this case, the effects of noise are found to be significant only at high noise levels and for large enough correlations times. [ABSTRACT FROM AUTHOR]

Possibilities of obtaining the analytical solutions of the one-dimensional equation of radiative transfer in resonance atomic/ionic lines in the Biberman-Holstein model for testing the edge plasma codes are investigated. It is shown that for some types of similarity of spatial profiles of three characteristics, namely, background plasma density, line shape width and non-radiation source of atomic excitation, the profile of excited atoms density may be described analytically in terms of the similarity of the above-mentioned profiles. (© 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim) [ABSTRACT FROM AUTHOR]

Implementation of self-consistent model of plasma recombination into the BIT1 PIC code and the simulation of detached SOL plasma are described. Our simulations indicate that in a strongly recombining plasma edge the sheath properties do not change qualitatively. The most affected parameter is the sheath heat transmission coefficient, which can increase by order of magnitude. [ABSTRACT FROM AUTHOR]

Five ELMy H-mode Ne seeded JET pulses have been simulated with the self-consistent core-SOL model COREDIV. In this five pulse series only the Ne seeding rate was changed shot by shot, allowing a thorough study of the effect of Ne seeding on the total radiated power and of its distribution between core and SOL tobe made. The increase in the simulations of the Ne seeding rate level above that achieved in experiments shows saturation of the total radiated power at a relatively low radiated-heating power ratio ( f rad = 0.60) and a further increase of the ratio of SOL to core radiation, in agreement with the reduction of W release at high Ne seeding level. In spite of the uncertainties caused by the simplified SOL model of COREDIV (neutral model, absence of ELMs and slab model for the SOL), the increase of the perpendicular transport in the SOL with increasing Ne seeding rate, which allows to reproduce numerically the experimental distribution core-SOL of the radiated power, appears to be of general applicability. [ABSTRACT FROM AUTHOR]

There is experimental evidence that the pedestal dynamics in type-I ELMy H-mode discharges is significantly affected by a change in the recycling conditions at the tungsten plasma-facing components (W-PFCs) after an ELM event. The integrated code JINTRAC has been employed to assess the impact of recycling conditions during type-I ELMs in JET ITER-like wall H-mode discharges. By employing a heuristic approach, a model to mimic the physical processes leading to formation and release (i.e. outgassing) of finite near-surface fuel reservoirs in W-PFCs has been implemented into the EDGE2D-EIRENE plasma-wall interaction code being part of JINTRAC. As main result it is shown, that a delay in the density pedestal build-up after an ELM event can be provoked by reduced recycling induced by depleted W-PFC particle near-surface reservoirs. However the pedestal temperature evolution is barely affected by the change in recycling parameters suggesting that the presented model is incomplete. [ABSTRACT FROM AUTHOR]

In this paper we present the comparison of simulations with the numerical codes COREDIV and SOLPS5.0 for JET L-mode discharges with ITER like wall (ILW). The simulations have been performed for L-mode shots with and without nitrogen seeding (#82291 - 9) which are characterised by relatively low auxiliary heating power ( P NBI = 1.1 MW) and low electron density ( ne = 2.35 × 1019 m-3). Comparisons are made to the experimental measurements (e.g. radiation levels, plasma profiles) and the differences between the results from the two codes (e.g. temperature and density profiles at the outer divertor plate) are shown and discussed. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The self-consistent COREDIV code is used to simulate discharges in a tokamak plasma, especially the influence of impurities during nitrogen and argon seeding on the key plasma parameters. The calculations are performed with and without taking into account the W prompt redeposition in the divertor area and are compared to the experimental results acquired on ASDEX Upgrade tokamak (shots #29254 and #29257). For both impurities the modeling shows a better agreement with the experiment in the case without prompt redeposition. It is attributed to higher average tungsten concentration, which on the other hand seriously exceeds the experimental value. By turning the prompt redeposition process on, the W concentration is lowered, what, in turn, results in underestimation of the radiative power losses. By analyzing the influence of the transport coefficients on the radiative power loss and average W concentration it is concluded that the way to compromise the opposing tendencies is to include the edge-localized mode flushing mechanism into the code, which dominates the experimental particle and energy balance. Also performing the calculations with both anomalous and neoclassical diffusion transport mechanisms included is suggested. [ABSTRACT FROM AUTHOR]

To facilitate the design of the China Fusion Engineering Testing Reactor (CFETR), predictive modeling for the assessment and optimization of the divertor performances is an indispensable approach. This paper presents the modeling of the edge plasma behaviors as well as the W erosion and transport properties in CFETR with ITER-like divertor by using the B2-Eirene/SOLPS 5.0 code package together with the Monte Carlo impurity transport code DIVIMP. As expected, SOLPS modeling of divertor-SOL plasmas finds that the peak heat flux onto the divertor targets greatly exceeds 10 MW/m, an engineering limit posed to the steady-state and/or long-pulse operation of the next-step fusion devices, for a wide range of plasma conditions, and thus modeling of Ar puffing by scanning the puffing rate for radiative divertor is performed. As the increase of the Ar puffing rate, the peak target heat fluxes and plasma temperature decreases exponentially,reflecting that Ar puffing is highly effective at power exhausting. Based on the ion fluxes from SOLPS, the W erosion is calculated by taking into consideration the bombardment of both D and Ar ions, and then the W plasma concentrations are calculated based on the W erosion fluxes using DIVIMP. The calculations show that if the Ar puffing only being used to reduce the divertor heat load, the W plasma contamination in the core plasma exceeds the tolerable value (<10), which demonstrates that some further upgrading of the divertor geometry is still needed. [ABSTRACT FROM AUTHOR]