ITER Similar Shape Research Articles

Compatibility of divertor detachment and ELM suppression in DIII-D high- βp plasmas with ITER-similar shape

Integration of transient and steady-state divertor heat fluxes control with a high-performance core is necessary for future fusion reactors. In recent DIII-D high- βp experiments, divertor detachment and simultaneous edge localized mode (ELM) suppression are demonstrated while the plasma confinement quality is maintained high in ITER-similar shape. By optimizing the neon injection in high- βp scenario with ITER-similar shape, deep detachment and ELM suppression are achieved with a high-performance core ( βN ∼ 2.8, βp ∼ 2.3) at q95 ∼ 7.5. Partial divertor detachment and suppression of large ELMs are achieved at q95 ∼ 6. The stability analyses suggest that with low neon injection, the density pedestal becomes higher and steeper and the Ti profile also increases, therefore the increased edge pressure and higher current density destabilize the Peeling-Ballooning mode (PBM), which would lead to a large ELM collapse. With strong neon gas puffing, the significantly reduced pedestal pressure and current density, due to the degraded Te pedestal, lead to the stabilization of PBM and ELMs are suppressed. For both cases, the coupling between the large radius internal transport barrier (ITB) and edge pedestal is the key reason for maintaining high global performance. The formation of large radius ITB compensates for pedestal degradation. Such results could provide an attractive scenario to well control the transient and steady-state heat flux onto the divertor plates while maintaining good plasma performance, which is an important step toward the steady-state operation of future fusion reactors.

Enhancing predictive capabilities in fusion burning plasmas through surrogate-based optimization in core transport solvers

This work presents the PORTALS framework (Rodriguez-Fernandez et al 2022 Nucl. Fusion 62 076036), which leverages surrogate modeling and optimization techniques to enable the prediction of core plasma profiles and performance with nonlinear gyrokinetic simulations at significantly reduced cost, with no loss of accuracy. The efficiency of PORTALS is benchmarked against standard methods, and its full potential is demonstrated on a unique, simultaneous 5-channel (electron temperature, ion temperature, electron density, impurity density and angular rotation) prediction of steady-state profiles in a DIII-D ITER Similar Shape plasma with GPU-accelerated, nonlinear CGYRO (Candy et al 2016 J. Comput. Phys. 324 73–93). This paper also provides general guidelines for accurate performance predictions in burning plasmas and the impact of transport modeling in fusion pilot plants studies.

Simultaneous reproduction of experimental profiles, fluxes, transport coefficients, and turbulence characteristics via nonlinear gyrokinetic profile predictions in a DIII-D ITER similar shape plasma

Experimental conditions obtained on the DIII-D tokamak in the ITER Similar Shape (ISS) have been compared extensively with nonlinear gyrokinetic simulation using the CGYRO code [Candy et al., J. Comput. Phys. 324, 73–93 (2016)] with comparisons spanning ion and electron heat fluxes, electron and impurity particle transport, and turbulent fluctuation levels and characteristics. Bayesian optimization techniques [Rodriguez-Fernandez et al., Nucl. Fusion 62(7), 076036 (2022)], combined with nonlinear gyrokinetics, have been used to obtain simultaneously Qi, Qe, and Γe flux-matched profiles that are found to be in good agreement with experimental profile measurements. Synthetic diagnostics were used to compare measured beam emission spectroscopy and correlation electron cyclotron emission turbulent fluctuations with nonlinear simulation. Although some disagreements exist, nonlinear simulations are found to be in generally good agreement with measured fluctuation levels, spectral shapes, and measured radial trends in low-k δne/ne and δTe/Te. Low (Li and C) and mid-Z (Ca) impurity transport was also compared with these flux-matched simulations. Fully stripped, low-Z impurities are well reproduced by the gyrokinetic modeling while clear disagreement exists in comparisons with mid-Z impurities. Nonlinear gyrokinetic investigation into the Z dependence of impurity transport in the ISS conditions is also performed, demonstrating clear trends of impurity diffusion with impurity Z (both D∝Z and D∝1/Z) that vary with the radial location studied. These trends are shown to result from the local dominance of ion temperature gradient or ∇n driven trapped electron mode turbulence and may contribute to the disagreement between simulation and experiment in mid-Z impurity transport. The results of this work represent one of the most complete validation studies of the gyrokinetic model performed to date and provide an example of new capabilities for predicting performance in future fusion devices.

Reducing the L-H transition power threshold in ITER-similar-shape DIII-D hydrogen plasmas

Recent dedicated DIII-D experiments in low-torque, ITER-similar-shape (ISS) hydrogen plasmas (at a plasma current I p ∼ 1.5 MA and ITER-similar edge safety factor q 95 ∼ 3.6) show that the L-H transition power threshold P LH can be reduced substantially (∼30%) with L-mode helium admixtures n He/n e ⩽ 25%. In the ensuing H-mode, helium ion fractions n He/n H remain below 25%. H-mode normalized pressure and confinement quality are only slightly affected by helium seeding, and Z eff ⩽ 2.15 (including helium and carbon content). The plasmas investigated here are electron-heat dominated, with temperatures T e(0)/T i(0) ⩾ 1 and edge heat flux ratio Q e/Q i(ρ = 0.95) ∼ 1.2–1.5. Without mitigation, P LH is higher by a factor of 2–3 in comparison to similar ISS deuterium plasmas. ISS hydrogen plasmas with lower plasma current I p ∼ 1 MA (increased edge safety factor q 95 ∼ 5.1) exhibit a substantially lower power threshold. This plasma current dependence, also observed previously on ASDEX-U and in JET, is not accounted for by the commonly used 2008 ITPA multi-machine threshold scaling, but could potentially allow H-mode access at marginal heating power during the initial plasma current ramp-up. Attempts to reduce P LH with low-field- and high-field-side hydrogen pellet injection, using 1.7 mm diameter pellets, have not demonstrated a robust threshold reduction, in contrast to successful earlier experiments with larger 2.7 mm pellets. Techniques for reducing P LH are very important for ITER, in particular for accessing H-mode in hydrogen plasmas during the Pre-Fusion Power Operation-1 (PFPO-1) campaign with marginal auxiliary heating power (20–30 MW of ECH).

Exploration of RMP ELM control on ITER-similar shape (ISS) in KSTAR

Recent control advances at KSTAR enabled us to not only establish but also to sustain the ITER-similar shape (ISS) in a stationary manner, matching the ITER-like dimensionless shape parameters in lower single null configuration with vastly contrasting upper/lower triangularities. The kinetic parameters of the ISS are different from the typical KSTAR configurations seen in ELM-RMP suppression experiments. We experimentally demonstrate the ISS-compatible resonant magnetic perturbation (RMP) control in KSTAR for the first time, using the n = 2, +90 degrees phasing RMP.

Evolution of ELMs, pedestal profiles and fluctuations in the inter-ELM period in NBI- and ECH-dominated discharges in DIII-D

In DIII-D, it has been observed that ELM frequency decreases by 40% and ELM spacing becomes more regular in time when heating is changed from pure neutral beam injection (NBI) to predominantly electron cyclotron heating (ECH) in ITER similar shape plasmas. In comparison with the pure NBI discharges, pedestal fluctuations in magnetics and density increase in the ECH-dominated discharges. Recovery of the pedestal profiles like electron density (n e), temperature (T e) and pressure (p e) shows marked differences for these two heating schemes. Average profiles in the last 30% of the ELM cycle show higher T e, lower n e, and similar p e at the pedestal top for the ECH discharge when compared to the NBI discharge. The gradient of T e (∇T e) is also steeper at the pedestal in the ECH discharge. Magnetic fluctuations show three distinct modes at 13–116 kHz in the ECH discharges only. n e fluctuations show two modes evolving in the inter-ELM period of the ECH discharge, a low-frequency (400 kHz) quasi-coherent mode (LFQC) and high-frequency (∼2 MHz) broadband (HFB) fluctuations. Evolution of these modes has marked correspondence with the inter-ELM ∇T e recovery. A sharp decrease in the D α baseline is observed whenever the LFQC weakens and the HFB grows, prior to each large ELM. Transport coefficients obtained from TRANSP show that MTM and/or TEM are plausible candidates for the observed fluctuations. Linear gyrofluid simulation (TGLF) corroborates this characterization. TGLF shows that the linear growth rate of the most dominant mode peaks at ion-scale (k θ ρ s ∼ 0.4) at the pedestal steep gradient and the frequency is in the electron diamagnetic direction. It is proposed that increased fluctuations in the ECH-dominated case, due to increased ∇T e, caused an increase in fluctuation-driven transport in the pedestal and slowed the pedestal recovery between ELMs, leading to a reduction in the ELM frequency.

Wide Operational Windows of Edge-Localized Mode Suppression by Resonant Magnetic Perturbations in the DIII-D Tokamak.

Edge-localized mode (ELM) suppression by resonant magnetic perturbations (RMPs) generally occurs over very narrow ranges of the plasma current (or magnetic safety factor q_{95}) in the DIII-D tokamak. However, wide q_{95} ranges of ELM suppression are needed for the safety and operational flexibility of ITER and future reactors. In DIII-D ITER similar shape plasmas with n=3 RMPs, the range of q_{95} for ELM suppression is found to increase with decreasing electron density. Nonlinear two-fluid MHD simulations reproduce the observed q_{95} windows of ELM suppression and the dependence on plasma density, based on the conditions for resonant field penetration at the top of the pedestal. When the RMP amplitude is close to the threshold for resonant field penetration, only narrow isolated magnetic islands form near the top of the pedestal, leading to narrow q_{95} windows of ELM suppression. However, as the threshold for field penetration decreases with decreasing density, resonant field penetration can take place over a wider range of q_{95}. For sufficiently low density (penetration threshold) multiple magnetic islands form near the top of the pedestal giving rise to continuous q_{95} windows of ELM suppression. The model predicts that wide q_{95} windows of ELM suppression can be achieved at substantially higher pedestal pressure in DIII-D by shifting to higher toroidal mode number (n=4) RMPs.

The role of edge resonant magnetic perturbations in edge-localized-mode suppression and density pump-out in low-collisionality DIII-D plasmas

Two-fluid nonlinear MHD simulations using the TM1 code demonstrate that the formation of magnetic islands at the top and bottom of the H-mode pedestal, together with the strong screening of resonant fields in the gradient region of the pedestal, can account for ELM suppression and density pump-out by n = 2 resonant magnetic perturbations (RMPs) in low-collisionality DIII-D ITER similar shape (ISS) plasmas. Using experimentally relevant transport coefficients, neoclassical resistivity, electron collisionality, and RMP amplitudes, nonlinear MHD simulations reproduce the observed level of density reduction (density pump-out) in DIII-D due to the formation of narrow magnetic islands in the resistive foot of pedestal. For large amplitude RMPs (Br/Bt > 1×10−4) simulations predict resonant field penetration and resulting in pressure reduction at the top of the pedestal, consistent with experimental observations at the onset of ELM suppression. The predicted reduction in the height and width of the pedestal by magnetic island enhanced collisional transport provides a quantitative mechanism for the stabilization of the peeling-ballooning mode (PBM). Importantly, these simulations predict strong screening of resonant fields in the steep gradient region of the pedestal due to strong E×B rotation and diamagnetic flows. However, if the plasma resistivity is made artificially larger (∼ 10X) than neoclassical, the simulations predict magnetic stochasticity throughout the plasma edge and the collapse of the pedestal due to the reduction in the penetration threshold with increasing resistivity. A scaling relation for resonant field penetration at the pedestal top, using several hundred nonlinear simulations, reproduces the density and E×B rotation dependence of the ELM suppression threshold observed in DIII-D. Simulations using ITER model equilibria suggest that the penetration threshold at the top of the ITER pedestal will be lower than in DIII-D due to the anticipated lower perpendicular flow velocities in ITER, resulting in the weaker screening of resonant fields.

The density dependence of edge-localized-mode suppression and pump-out by resonant magnetic perturbations in the DIII-D tokamak

The density dependence of edge-localized-mode (ELM) suppression and density pump-out (density reduction) by n = 2 resonant magnetic perturbations (RMPs) is consistent with the effects of narrow well-separated magnetic islands at the top and bottom of the H-mode pedestal in DIII-D low-collisionality plasmas. Nonlinear two-fluid MHD simulations for DIII-D ITER similar shape discharges show that, at low collisionality (ν*e < 0.5), low pedestal density is required for resonant field penetration at the pedestal top (ne,ped ≈ 2.5 × 1019 m−3 at ψN ≈ 0.93), consistent with the ubiquitous low density requirement for ELM suppression in these DIII-D plasmas. The simulations predict a drop in the pedestal pressure due to parallel transport across these narrow width (ΔψN ≈ 0.02) magnetic islands at the top of the pedestal that is stabilizing to Peeling-Ballooning-Modes and comparable to the pedestal pressure reduction observed in experiment at the onset of ELM suppression. The simulations predict density pump-out at experimentally relevant levels (Δne/ne ≈ −20%) at low pedestal collisionality (ν*e ≈ 0.1) due to very narrow (ΔψN ≈ 0.01–0.02) RMP driven magnetic islands at the pedestal foot at ψN ≈ 0.99. The simulations show decreasing pump-out with increasing density, consistent with experiment, resulting from the inverse dependence of parallel particle transport on collisionality at the foot of the pedestal. The robust screening of resonant fields is predicted between the top and bottom of the pedestal during density pump-out and ELM suppression, consistent with the preservation of strong temperature gradients in the edge transport barrier as seen in experiment.

Advances in the steady-state hybrid regime in DIII-D—a fully non-inductive, ELM-suppressed scenario for ITER

The hybrid regime with beta, collisionality, safety factor and plasma shape relevant to the ITER steady-state mission has been successfully integrated with ELM suppression by applying an odd parity n = 3 resonant magnetic perturbation (RMP). Fully non-inductive hybrids in the DIII-D tokamak with high beta ( ⩽ 2.8%) and high confinement (H98y2 ⩽ 1.4) in the ITER similar shape have achieved zero surface loop voltage for up to two current relaxation times using efficient central current drive from ECCD and NBCD. The n = 3 RMP causes surprisingly little increase in thermal transport during ELM suppression. Poloidal magnetic flux pumping in hybrid plasmas maintains q above 1 without loss of current drive efficiency, except that experiments show that extremely peaked ECCD profiles can create sawteeth. During ECCD, Alfvén eigenmode (AE) activity is replaced by a more benign fishbone-like mode, reducing anomalous beam ion diffusion by a factor of 2. While the electron and ion thermal diffusivities substantially increase with higher ECCD power, the loss of confinement can be offset by the decreased fast ion transport resulting from AE suppression. Extrapolations from DIII-D along a dimensionless parameter scaling path as well as those using self-consistent theory-based modeling show that these ELM-suppressed, fully non-inductive hybrids can achieve the Qfus = 5 ITER steady-state mission.

The energy confinement response of DIII-D plasmas to resonant magnetic perturbations

Resonant magnetic perturbations (RMPs) are a leading method for edge localized modes (ELMs) Control in fusion plasmas. However they can also cause a rapid degradation in energy confinement. In this paper we show that the energy confinement in low collisionality ( < 0.3) DIII-D ITER similar shape (ISS) plasmas often recovers after several energy confinement times for RMP amplitudes up to the threshold for ELM suppression. Immediately following the application of the RMP, the plasma stored energy decreases in proportion to the decrease in the line-averaged density during density ‘pump-out’. Later in the discharge confinement recovery is observed in the thermal ion channel and is correlated with the increase in the ion temperature at the top of the H-mode pedestal. A correlation between the inverse scale length of the ion temperature () and the shearing rate at the top of the pedestal is seen during the confinement recovery phase. Transport analysis reveals that the confinement improvement in the ion channel results from the self-similarity in the ion temperature profiles in the plasma core combined with the observed increase in in the plasma edge following density pump-out. In contrast the electron temperature scale length () remains essentially unchanged in response to the application of the RMP. At significantly higher RMP levels the edge shearing rate and does not increase and the confinement does not recover following density pump-out.

Effect of rotation zero-crossing on single-fluid plasma response to three-dimensional magnetic perturbations

In order to understand the effect of rotation on the response of a plasma to three-dimensional magnetic perturbations, we perform a systematic scan of the zero-crossing of the rotation profile in a DIII-D ITER-similar shape equilibrium using linear, time-independent modeling with the M3D-C1 extended magnetohydrodynamics code. We confirm that the local resonant magnetic field generally increases as the rotation decreases at a rational surface. Multiple peaks in the resonant field are observed near rational surfaces, however, and the maximum resonant field does not always correspond to zero rotation at the surface. Furthermore, we show that non-resonant current can be driven at zero-crossings not aligned with rational surfaces if there is sufficient shear in the rotation profile there, leading to amplification of near-resonant Fourier harmonics of the perturbed magnetic field and a decrease in the far-off-resonant harmonics. The quasilinear electromagnetic torque induced by this non-resonant plasma response provides drive to flatten the rotation, possibly allowing for increased transport in the pedestal by the destabilization of turbulent modes. In addition, this torque acts to drive the rotation zero-crossing to dynamically stable points near rational surfaces, which would allow for increased resonant penetration. By one or both of these mechanisms, this torque may play an important role in bifurcations into suppression of edge-localized modes. Finally, we discuss how these changes to the plasma response could be detected by tokamak diagnostics. In particular, we show that the changes to the resonant field discussed here have a significant impact on the external perturbed magnetic field, which should be observable by magnetic sensors on the high-field side of tokamaks but not on the low-field side. In addition, TRIP3D-MAFOT simulations show that none of the changes to the plasma response described here substantially affects the divertor footprint structure.

Long-pulse stability limits of the ITER baseline scenario

DIII-D has made significant progress in developing the techniques required to operate ITER, and in understanding their impact on performance when integrated into operational scenarios at ITER-relevant parameters. Long duration plasmas, stable to m/n = 2/1 tearing modes, with an ITER-similar shape and Ip/aBT, have been demonstrated in DIII-D, that evolve to stationary conditions. The operating region most likely to reach stable conditions has normalized pressure, βN ≈ 1.9–2.1 (compared to the ITER baseline design of 1.6–1.8), and a Greenwald normalized density fraction, fGW 0.42–0.70 (the ITER design is fGW ≈ 0.8). The evolution of the current profile, using internal inductance (li) as an indicator, is found to produce a smaller fraction of stable pulses when li is increased above ≈1.1 at the beginning of βN flattop. Stable discharges with co-neutral beam injection are generally accompanied with a benign n = 2 magnetohydrodynamic mode. However if this mode exceeds ≈10 G, the onset of a m/n = 2/1 tearing mode occurs with a loss of confinement. In addition, stable operation with low applied external torque, at or below the extrapolated value expected for ITER has also been demonstrated. With electron cyclotron injection, the operating region of stable discharges has been further extended at ITER equivalent levels of torque and to edge-localized mode (ELM) free discharges at higher torque but with the addition of an n = 3 magnetic perturbation from the DIII-D internal coil set. The characterization of the ITER baseline scenario evolution for long pulse duration, extension to more ITER-relevant values of torque and electron heating, and suppression of ELMs have significantly advanced the physics basis of this scenario, although significant effort remains in the simultaneous integration of all these requirements.

Striation pattern of target particle and heat fluxes in three dimensional simulations for DIII-D

The application of resonant magnetic perturbations results in a non-axisymmetric striation pattern of magnetic field lines from the plasma interior which intersect the divertor targets. The impact on related particle and heat fluxes is investigated by three dimensional computer simulations for two different recycling conditions (controlled via neutral gas pumping). It is demonstrated that a mismatch between the particle and heat flux striation pattern (splitting vs. no splitting), as is repeatedly observed in ITER similar shape H-mode plasmas at DIII-D, can be reproduced by the simulations for high recycling conditions at the onset of partial detachment. These results indicate that a detailed knowledge of the particle and energy balance is at least as important for realistic simulations as the consideration of a change in the magnetic field structure by plasma response effects.

Impact of screening of resonant magnetic perturbations in three dimensional edge plasma transport simulations for DIII-D

The impact of resonant magnetic perturbations (RMPs) on the plasma edge can be analyzed in detail by three dimensional computer simulations, which take the underlying magnetic field structure as input. Previously, the “vacuum approximation” has been used to calculate the magnetic field structure although plasma response effects may result in a screening (or even an amplification) of the external perturbations. Simulation results for an ITER similar shape plasma at the DIII-D tokamak are presented for the full vacuum perturbation field and an ad hoc screening case in comparison to the unperturbed configuration. It is shown that the RMP induced helical patterns in the plasma edge and on the divertor target shrink once screening is taken into account. However, a flat temperature profile is still found in the “open field line domain” inside the separatrix, while the “density pump out effect” found in the vacuum RMP case is considerably weakened.

On gas flow effects in 3D edge transport simulations for DIII-D plasmas with resonant magnetic perturbations

Pumping of neutral gas and re-fuelling in other locations (e.g. neutral beam injection or gas puffing) are key ingredients in divertor physics. However, these have not been included in previous 3D edge plasma transport simulations with the EMC3–EIRENE code. Including these effects brings the simulations closer to reality and we demonstrate that this has a significant impact on divertor parameters. In particular, we study the impact of the pumping efficiency on ITER similar shape plasmas at the DIII-D tokamak in the presence of resonant magnetic perturbations. Further, we investigate the so-called ‘particle pump-out effect’. We show that in the present transport model and for given pumping and re-fuelling rates, the density at the transition to the core region decreases with increasing perturbation current and also the temperature. The striation patterns in the target particle and heat fluxes are extended into the regular scrape-off layer, which is caused by cross-field transport into a thin layer around the perturbed separatrix.

Changes in density fluctuations as a result of resonant magnetic perturbations correlate with the density inverse scale length

In this paper we show that the changes in ion-scale density fluctuations ñ/n induced by resonant magnetic perturbations (RMPs) exhibit the same trends as modifications to the inverse density scale length, a/Ln in the pedestal area. In a series of ITER similar shape (ISS) H-mode discharges in DIII-D, the n = 3 even parity RMP-coil current is varied from 4 kA up to 6.2 kA. The application of n = 3 RMPs results in an increase of a/Ln in the plasma core and a reduction in the pedestal area. Comparing the changes in a/Ln with the changes in density ion-scale fluctuations ñ/n shows that at ΨN=0.95, the ñ/n follows the trends in a/Ln. This suggests that the changes in ñ/n are due to changes in the underlying density gradient rather than due to the RMPs directly in ISS H-mode discharges in DIII-D.

Simulating ITER plasma startup and rampdown scenarios in the DIII-D tokamak

DIII-D experiments have investigated ITER startup scenarios, including an initial phase where the plasma was limited on low field side poloidal bumper limiters. In addition, li feedback control has been tested with the goal of producing discharges in ITER within the capabilities of the poloidal field coil set and favourable to the intended mode of operations in the subsequent constant current (flattop) phase. These discharges have been modelled using the Corsica free boundary equilibrium code. High performance hybrid scenario discharges (βN = 2.8, H98,y2 = 1.4) and ITER H-mode baseline discharges (βN > 1.6, H98,y2 = 1–1.2) have been obtained experimentally in an ITER similar shape after the ITER-relevant startup. Studies have been initiated to develop a reliable scenario for exiting the burn phase and ramping down the plasma current in ITER without disruptions.

Comparison of edge plasma perturbation during ELM control using one vs. two toroidal rows of RMP coils in ITER similar shaped plasmas on DIII-D

Large Type-I edge localized modes (ELMs) were suppressed by n = 3 resonant magnetic perturbations (RMPs) from a set of internal coils in plasmas with an ITER similar shape at the ITER pedestal collisionality, ν e ∗ ∼ 0.1 and low edge safety factor ( q 95 ≈ 3.6), with either a single toroidal row of the internal RMP coils or two poloidally separated rows of coils. ELM suppression with a single row of internal coils was achieved at approximately the same q 95 surface-averaged perturbation field as with two rows of coils, but required higher current per coil. Maintaining complete suppression of ELMs using n = 3 RMPs from a single toroidal row of internal coils was less robust to variations in input neutral beam injection torque than previous ELM suppression cases using both rows of internal coils. With either configuration of RMP coils, maximum ELM size is correlated with the width of the edge region having good overlap of the magnetic islands from vacuum field calculations.

Aspects of three dimensional transport for ELM control experiments in ITER-similar shape plasmas at low collisionality in DIII-D

A study of three-dimensional (3D) perturbed magnetic field structures and transport for edge localized mode control experiments with resonant magnetic perturbations at DIII-D is presented. We focus on ITER-Similar Shape plasmas at ITER relevant electron pedestal collisionalities . This study is performed in comparison with results from TEXTOR-Dynamic Ergodic Divertor circular limiter plasmas. For both experiments the magnetic field structure is analyzed in the vacuum paradigm—superimposing the external RMP field on the unperturbed equilibrium. For TEXTOR L-mode plasmas this description holds for normalized poloidal flux ΨN > 0.7 without tearing modes driven by the RMP field. For DIII-D H-mode plasmas the validity of this approach still needs to be established. In this paper a method is discussed to diagnose the degree of edge stochastization based on a comparison between modeled magnetic footprints on the divertor targets and experimental data. Clear evidence is presented for the existence of a generic separatrix perturbation causing striation of target particle fluxes. However, heat fluxes into these striations are small. This observation can be explained by accounting for the different heat and particle source locations and the 3D trajectories of the open, perturbed field lines toward the divertor target. Analysis of the transport characteristics filling the perturbed separatrix lobes based on initial EMC3/EIRENE modeling suggests the existence of open field lines connecting the stochastic edge to the target pattern. However, the width and inward most extent of the actual stochastic layer cannot yet be quantified.