Divertor Plasma Detachment Research Articles

The role of divertor pumping combined with full drifts in particle exhaust and divertor plasma

The effect of drifts combined with pumping on particle exhaust is assessed using the SOLPS-ITER code package, considering full drifts. Both drift and pumping speed S can affect particle exhaust. Drifts change the neutral density by influencing plasma flow and the resulting particle recycling. This leads to the accumulation of neutral particles either far away or close to the pump opening location. The particle exhaust is enhanced as S rises. When the pump opening is positioned at the common flux region (CFR) of the outer divertor (referred to as Pump CFR/OD), particle exhaust is suppressed by drifts in forward Bt , while it is enhanced by drifts in reversed Bt , with fixed S. On the other hand, when the pump is situated in the private flux region (PFR) of the OD (referred to as Pump PFR/OD), particle exhaust is enhanced by drifts in both reversed and forward Bt compared to the case without drifts. Moreover, the effective pumping in reversed Bt is stronger than in forward Bt . In the same Bt direction, Pump PFR/OD has a higher effective pumping than Pump CFR/OD. Increased S results in higher particle exhaust in all Bt direction and pump location cases. The plasma detachment is affected by drift, S and pump opening location, respectively. For the specified Bt direction and pump opening location case, higher S suppresses plasma detachment. For identical particle exhaust rates, stronger pumping capacity can promote plasma detachment. Therefore, Pump PFR/OD can more easily achieve OD detachment than Pump CFR/OD in the same Bt direction. Overall, placing the pump at the PFR side of the OD while running in reversed Bt is the best option from the divertor particle exhaust and plasma detachment point of view.

Performance assessment of spectroscopic measurements of hydrogen and beryllium influxes and ionization fronts in ITER divertor plasmas using the divertor impurity monitor

Characteristics of Hα and Be II emissions in ITER divertor plasmas have been evaluated, and the applicability of a spectroscopic evaluation method of influxes and ionization front positions by using the divertor impurity monitor (DIM) has been systematically verified for the first time based on plasma profile datasets with parameter scans. In addition, practical values of the ionization/photon factor (S/XB) for DIM to evaluate influxes have been determined. In attached divertor plasmas and the onset of plasma detachment, H influx integrated over the entire area of the divertor can be evaluated within ±30 % accuracy using Hα intensities measured by the DIM and S/XB value of 32, which is higher than the S/XB in most divertor plasmas in previous devices owing to high electron density. The ionization front position can also be evaluated to within a few mm by using peak channel position. However, when electron temperature further decreases from the detachment onset, the ionizing plasma component (caused by electron impact excitation) in the total Hα emission decreases and the spatial profiles of emission and ionization diverge, making spectroscopic evaluation difficult. In all conditions, Hα intensities could be measured with a more than sufficient number of photons at the required time resolution using the current DIM design. For Be influx integrated over the entire divertor area, it has been found that Be I is not suitable for the DIM since its intensity is similar to the background continuous spectra. Replacing Be I with Be II solves this problem. Even in detached divertor plasmas, it is possible to evaluate Be1+ ionization flux within ±30 % accuracy using Be II intensities and S/XB, the value of which depends on the DIM’s fields of view. However, the evaluated Be1+ ionization flux is lower than Be influx by more than 80 %, meaning that the evaluation of Be influx using Be II would be an underestimate. But, Be II can be measured with a more than sufficient number of photons.

Development of Monte Carlo neutral transport code and the application to methane seeding plasma detachment

AbstractThe collisions between neutral particles and plasma in the divertor region determine divertor detachment and heat flux to the target. However, the diagnostic of neutral particle information in the tokamak experiment is very lacking. To study the behavior of neutral particles in the boundary region flexibly and comprehensively, a two‐dimensional kinetic neutral transport code based on Monte Carlo method is developed in this work. The code is applied to simulate the neutral particles in the boundary of EAST device. Firstly, the simulation results are compared with EIRENE using an identical plasma background, and good agreement is obtained. Secondly, the transport of methane and the corresponding hydrocarbon compounds is simulated using the developed code. The methane injection case is compared with the injection of carbon and deuterium molecule mixture gas (ratio 1:2) to investigate their impact on detachment and fueling. The simulation reveals that, compared to the direct carbon atom injection case, charge‐exchange collision is the dominant collision of neutral methane (hydrocarbons), which suppresses divertor power radiation and plasma detachment while promoting the improvement of fueling efficiency.

UEDGE modeling of plasma detachment of CFETR with ITER‐like divertor geometry by external impurity seeding

AbstractEfficient handling of high heat flux on the plasma‐facing components, particularly the divertor targets, poses a significant challenge for the Chinese Fusion Engineering Testing Reactor (CFETR) with fusion power of Gigawatt. This work investigates the divertor plasma detachment of CFETR with a standard ITER‐like divertor geometry by neon (Ne) or argon (Ar) impurity seeding using UEDGE code. The cross‐field drifts terms are switched off, and fluid neutral models and a “fixed‐fraction” impurity model are applied to enable efficient simulations for the study of CFETR detachment. In order to reduce the heat load on the divertor targets below the acceptable level (<10 MW/m2), the impurity fraction (f), pumping speed (S), and upstream density are varied to identify the suitable operations window during Ne seeding. The effects of Ne and Ar impurities on the plasma detachment are compared. It is found that with the power across the core‐edge interface PSOL = 200 MW and separatrix density of 2.8 1019 , Ne impurity fraction ≥1.7%, and Ar impurity fraction ≥0.24% can achieve the partial detachment. Achieving similar total radiation power (˜148 MW), the Ne fraction is 2.3% and the Ar fraction is 0.24%. Moreover, the simulation results indicate that Ar exhibits better power radiation efficiency and core compatibility compared with Ne.

Recent progress of plasma exhaust concepts and divertor designs for tokamak DEMO reactors

The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2.Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

Study on ELM burn-through induced by small ELMs with divertor detachment in EAST

In the H-mode radiative divertor experiments, the transient heat flux induced by edge localized modes (ELMs) could cause the detached divertor plasma to re-attach in a few milliseconds, which is so-called ELM burn-through. The periodic hot plasma burning through the neutral front could bring significant damages to the divertor target and thus has attracted increasing attention. A systematic study on the burn-through of small ELMs during radiative divertor experiments has been conducted in the EAST tokamak. Firstly, the affected area of small ELMs on the divertor target has been evaluated, and the results show that the radial extension of small ELMs is restricted to a range of less than the particle flux fall-off width (λjs) along the divertor target. Secondly, it has been found that the occurrence of small ELM burn-through is highly correlated with the degree of divertor detachment and ELM frequency. When the detached area is larger than λjs, the small ELMs in the main plasma could be mitigated or even suppressed, avoiding the burn-through of small ELMs. When the divertor detachment only occurs near the strike point while the ELM frequency is high enough (fELM ≳ 800 Hz), there is also no burn-through of small ELMs. When the degree of divertor detachment is low and ELM frequency is not high simultaneously, the ELM burn-through induced by small ELMs could be observed and is correlated with the competition between ionization and recombination process of the divertor plasma. Further analysis reveals that the burn-through is triggered by the accumulated particles and energy of multiple small ELMs, not by that of a single small ELM under this condition. Finally, statistical analysis indicates that the small ELM burn-through occurs when the detached area is limited to a region of less than 0.4λjs and simultaneously ELM frequency is smaller than 800 Hz.

Core integrated simulations for the Divertor Tokamak Test facility scenarios towards consistent core-pedestal-SOL modelling

Deuterium plasma discharges of the Divertor Tokamak Test facility (DTT) in different operational scenarios have been predicted by a comprehensive first-principle based integrated modelling activity using state-of-art quasi-linear transport models. The results of this work refer to the updated DTT configuration, which includes a device size optimisation (enlargement to m and a = 0.70 m) and upgrades in the heating systems. The focus of this paper is on the core modelling, but special attention was paid to the consistency with the scrape-off layer parameters required to achieve divertor plasma detachment. The compatibility of these physics-based predicted scenarios with the electromagnetic coil system capabilities was then verified. In addition, first estimates of DTT sawteeth and of DTT edge localised modes were achieved.

1D model of tokamak scrape-off layer and divertor plasmas: Dimensionless parameters and the results of numerical simulations

Solutions of a 1D model of the scrape-off layer and a divertor plasma in a tokamak are analyzed on the basis of dimensionless parameters governing the physics of both plasma and neutral transport and atomic physics processes. Numerical simulations based on fluid plasma and Monte Carlo neutral modeling confirm earlier results obtained with simplified analytic considerations and numerical simulations performed with fluid neutral description and demonstrate bifurcation of plasma parameters. These simulations also confirm that the major physical reasons for the rollover of the plasma flux to the target, which is the key feature of the phenomenon of divertor plasma detachment, are the impurity radiation loss and plasma recombination. It is suggested that simulations of the edge plasma based on dimensionless parameters provide a powerful tool for checking the consistency of the physics incorporated in existing complex 2D codes used for edge plasma modeling.

Fascinating physics at the edge of magnetic fusion devices

The physics of the processes at the edge of magnetic fusion devices is multifaceted and exhibits complex, nonlinear synergistic effects. Even though this region occupies only a small portion of the whole device, it plays a crucial role in overall plasma confinement, heat exhaust, and plasma–wall interactions. The latter affects not only the performance but also the lifetime of plasma-facing components and therefore remains an outstanding challenge for future fusion reactors. At the edge of fusion devices, researchers are dealing with phenomena including classical and anomalous plasma transport, atomic physics effects, and physics of plasma-facing material under strong irradiation by particle and energy fluxes. The diversity of edge physics makes it particularly attractive for young scientists. Working in this field, they can find endless possibilities to demonstrate their talents and creativity. This short review describes just some of the basic scrape-off layer and divertor plasma phenomena including divertor plasma detachment, intermittent bursts of anomalous cross-field plasma transport, plasma–material interactions, and dust in fusion plasmas, which are of particular interest in relation to fusion reactors.

Divertor plasma detachment: roles of plasma momentum, energy, and particle balances

Contrary to statements from some recent papers, the bifurcation of the scrape-off layer and divertor plasma parameters, as described within a 1D model, does exist. The analysis of both numerical simulations and experimental data shows that the real physical reasons that can cause a strong reduction of the plasma particle flux to the target are impurity radiation loss and volumetric plasma recombination. Fitting the parameter describing the ‘plasma momentum removal’, based on the results of 2D numerical simulations, then using this fit in subsequent analytic expressions for , and finally interpreting the evolution of as the result of elastic plasma-neutral interactions break the causality principle and lead to a wrong physical picture of divertor plasma detachment.

Radiation characteristics of detached divertor plasmas in W7-X

• Spatial distributions of line emissions from hydrocarbon and carbon have been measured in detached divertor plasma in W7-X. • Spatial distributions of line emissions from hydrogen molecules and atoms have been measured in detached divertor plasma in W7-X. • Hydrogen Balmer line ratios indicate that the detached plasma is in excitation or partial recombination state. • Electron pressure along flux tube in SOL drops significantly but not deeper due to the lack of volume recombination in detached divertor plasma. Detached divertor plasmas has been achieved by density ramp in W7-X. During the transition into detachment, the main divertor radiation region moves from the region close to the divertor targets to the region around LCFS. The typical spatial distributions of carbon and hydrogen Balmer lines in detached divertor plasma are presented. They show that the line emissions of CH produced by chemical sputtering is still strong close to the targets. The C I fills up the whole detached divertor plasma and its intensity increases when approaching the targets due to the dissociation from hydrocarbon and self-sputtering. The C II and C III are mainly located at around LCFS region. Meanwhile, the distribution of hydrogen Balmer lines is considerably flatter in detached divertor plasma. The Balmer line ratios indicate that the detached divertor plasma seems to be still in excitation or partial recombination state. This is consistent with the divertor Langmuir probe measurements which shows that the electron temperature near strike line is around 6 eV. Finally, the electron pressure at upstream and downstream position is measured and its ratio shows that the electron pressure drops significantly along the magnetic field lines. However, the pressure can not drop further due to the lack of the volume recombination in detached divertor plasma.

Data-driven model for divertor plasma detachment prediction

We present a fast and accurate data-driven surrogate model for divertor plasma detachment prediction leveraging the latent feature space concept in machine learning research. Our approach involves constructing and training two neural networks: an autoencoder that finds a proper latent space representation (LSR) of plasma state by compressing the multi-modal diagnostic measurements and a forward model using multi-layer perception (MLP) that projects a set of plasma control parameters to its corresponding LSR. By combining the forward model and the decoder network from autoencoder, this new data-driven surrogate model is able to predict a consistent set of diagnostic measurements based on a few plasma control parameters. In order to ensure that the crucial detachment physics is correctly captured, highly efficient 1D UEDGE model is used to generate training and validation data in this study. The benchmark between the data-driven surrogate model and UEDGE simulations shows that our surrogate model is capable of providing accurate detachment prediction (usually within a few per cent relative error margin) but with at least four orders of magnitude speed-up, indicating that performance-wise, it has the potential to facilitate integrated tokamak design and plasma control. Comparing with the widely used two-point model and/or two-point model formatting, the new data-driven model features additional detachment front prediction and can be easily extended to incorporate richer physics. This study demonstrates that the complicated divertor and scrape-off-layer plasma state has a low-dimensional representation in latent space. Understanding plasma dynamics in latent space and utilising this knowledge could open a new path for plasma control in magnetic fusion energy research.

Linear analysis of cross-field dynamics with feedback instability on detached divertor plasmas

A theoretical model of feedback instability is proposed to explain the mechanism of correlation between the detachment and the cross-field plasma transport. It is shown that the feedback instability on the detached divertor plasma can be induced in a certain condition in which the volume recombination frequency is larger than the ion cyclotron frequency in the recombination region. Further, the density gradient and the electric field in the direction perpendicular to the magnetic flux surface are not zero in the condition. The feedback instability can provide the cross-field plasma transport in the boundary layer of magnetic fusion torus devices. Furthermore, the properties of the radial transport observed in the NAGDIS-II linear device experiment are compared with the estimation by the feedback instability model. The dependence of the feedback instability mode on the total collision frequency and the recombination coefficient and the density gradient has been also investigated. Although the dependency on the total collision frequency and the recombination coefficient for the typical fusion torus device case is opposite to that for the NAGDIS-II case, the represented dependencies show the reasonable tendency in each case.

Volumetric recombination in EMC3-EIRENE: Implementation and first application to the pre-fusion power operation phase in ITER

An upgrade to the 3D plasma boundary model EMC3-EIRENE for detached divertor plasmas is introduced and evaluated. Stabilization of the iterative approximation of a self-consistent plasma–neutral gas solution is required at low divertor plasma temperatures, and this has been achieved by linearization of the energy loss term associated with the electron–neutral gas interaction. An implicit method for volumetric recombination has been implemented, which improves agreement with SOLPS-ITER results for the pre-fusion power operation phase in ITER. Resonant magnetic perturbations for control of edge localized modes are found to result in a partially detached divertor plasma with a significantly broader reduction of particle loads in the traditional strike zone, which is attributed to volumetric recombination. The additional non-axisymmetric far scrape-off layer strike points, on the other hand, remain attached at relatively high temperature so that volumetric recombination is not effective there.

Experimental study of the influence of gas puff locations on H-mode boundary plasmas with argon seeding on EAST

To investigate the optimal scenario of impurity seeding to obtain divertor plasma detachment for target protection, experiments with Ar&D2 seeding from two different poloidal locations, the upper outer (UO) divertor target and lower outer (LO) target, were carried out on EAST. Partial energy detachment (the electron temperature near the strike point T e,spt ⩽ 10 eV) were obtained with Ar&D2 mixture puffing from the UO target and, for the first time, from the LO target into H-mode plasmas in the upper single null (USN) configuration. The peak heat flux q t on the UO target was significantly reduced (by ∼80%). The rollover of ion flux density j s did not appear probably due to insufficient momentum loss, which is independent of the puff locations. The poloidal asymmetries of particle and heat fluxes on the targets have also been investigated. The UO-dominant asymmetry of particle flux was reversed, while the UO-dominant asymmetry of heat flux was mitigated but not reversed. The plasma confinement dropped by only 14% and 8.2% in the LO-puff case and UO-puff case, respectively, during detachment. The high level of C may contributed greatly to the higher radiation in the bulk plasma region and the greater decline in W MHD in the LO-puff case.

Observation of fully detached divertor integrated with improved core confinement for tokamak fusion plasmas

Integration of divertor detachment with a high-performance (βN ∼ 3, βp > 2, H98 ∼ 1.5) core plasma has been demonstrated in DIII-D high-βp (poloidal beta) plasmas associated with a sustained core internal transport barrier (ITB) and an H-mode edge transport barrier (ETB). Such good core-edge integration has been achieved for both neon and nitrogen seeding, for both favorable and unfavorable B-field directions, independently from the impurity puffing locations, though these variations play important roles on divertor characteristics. Compared to the standard H-mode plasmas, the high-βp plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te < 5 eV), low particle flux, and low heat flux across the entire divertor target plate were obtained by using nitrogen seeding. This detached high-βp plasma is compatible with a newly developed detachment control system which can help optimize the nitrogen gas flow rate. Several features, i.e., the high edge safety factor in the high-βp scenario, impurity injection, closed divertor and reduced heating power requirement due to the high confinement, facilitate the achievement of full divertor detachment at lower density. Instead of degrading global performance, the divertor detachment facilitates the access to an even stronger ITB at large radius with a relatively weak ETB through self-organized synergy between ITB and ETB, leading to sustained high confinement. The strengthening of the large-radius ITB compensates for the ETB degradation associated with divertor detachment. In addition, a weak ETB naturally has smaller edge localized modes (ELMs). In particular, with neon injection, a long-period no-ELM H-mode phase has been achieved simultaneously with high-performance core and partially detached divertor plasmas. These results demonstrate the possibility of integrating excellent core plasma performance with an effective divertor solution, an essential step toward steady-state operation of reactor-grade plasmas.

MHD stability constraints on divertor heat flux width in DIII-D

Abstract The radial width of the exhaust heat flux flowing in the SOL of DIII-D is found to expand at high input power and plasma density, consistent with MHD ballooning stability limits. At low heating power, ~3 MW, the SOL width remains constant and consistent with established empirical scaling dependent only on the midplane poloidal field. At high heating power, ~ 13 MW a higher separatrix density, and resulting higher separatrix pressure is required for divertor detachment. The separatrix pressure gradient at the separatrix continues to increase with density until it begins to saturate at levels ~50% above the calculated ideal MHD ballooning limit. Examination of the separate contributions to the pressure gradient from electrons and ions reveals the ion pressure gradient to saturate more strongly than the electron pressure gradient. Potential analysis issues leading to the measured pressure gradient exceeding the ballooning limit are discussed. At high density, particularly for detached divertor plasmas, the SOL width for temperature and density expand modestly, ~30–50%. The divertor plasma density profile in detachment also reflects this trend, expanding in the radial direction a factor of 2–3. Despite the SOL width expansion at the highest power and density no degradation of the pedestal and resulting core confinement is observed with the additional density at high power. These results imply a more favorable scaling for divertor heat flux control in future reactor-scale tokamaks than predicted by existing empirical scaling.

The role of drifts on the isotope effect on divertor plasma detachment in JET Ohmic discharges

• The magnitude of the isotope effect is determined by the pressure in the sub-divertor. • Pumping of gas within the JET divertor is effective near the outer target only. • The re-distribution of the neutrals due to drifts decreases the throughput. • The total decrease of the throughput reduces the isotope effect. Experiments in JET-ILW Ohmic confinement mode plasmas show that the line-averaged detachment onset density in deuterium discharges is approx. 10% lower than in hydrogen discharges. The magnitude of the isotope effect on the detachment onset density depends on the divertor geometry, the magnetic configuration and the throughput of the sub-divertor/divertor cryopump system. Simulations with the edge fluid code EDGE2D-EIRENE revealed that the pumping of neutral gas within the JET divertor is effective near the outer divertor target only. The studies show that the magnitude of the isotope effect is determined by the molecular pressure in the sub-divertor pumping plenum. According to the simulations, operating in vertical configurations or closer proximity of the strike point increases the molecular pressure (and thus throughput) in front of the outer pumping plenum by up to 15% compared to the horizontal configuration, thus producing a stronger isotope effect on the detachment onset density. Similarly, EDGE2D-EIRENE predicts that plasma, hence neutral, redistribution due to E × B drifts in favourable B T configurations (ion B × grad(B) towards the divertor) decreases the throughput. The total decrease of the throughput reduces the isotope effect on the detachment onset density, and decreases the detachment onset density for both isotopes.

Neutral impact on anomalous edge plasma transport and its correlation with divertor plasma detachment

The impact of neutrals on anomalous edge plasma transport and zonal flow (ZF) is considered. As an example, it is assumed that edge plasma turbulence is driven by the resistive drift wave (RDW) instability. It is found that the actual effect of neutrals is not related to a suppression of the instability per se, but due to an impact on the ZF. Particularly, it is shown that, whereas the neutrals make very little impact on the linear growth rate of the RDW instability, they can largely reduce the zonal flow generation in the non-linear stage, which results in an enhancement of the overall anomalous plasma transport. Even though only RDW instability is considered, it seems that such an impact of neutrals on anomalous edge plasma transport, i.e. neutrals enhance the transport via the reduction of ZF, has a very generic feature. It is conceivable that such neutral induced enhancement of anomalous plasma transport is observed experimentally in a detached divertor regime, which is accompanied by a strong increase of neutral density.

Influence of Cross-Field Transport in a Divertor on Seeded Impurity Radiation and Divertor Plasma Detachment

The dissipative divertor for the power and particle exhaust in a tokamak relies heavily on strong impurity radiation inside the divertor volume. In the present paper, the influence of the cross-field transport inside the divertor volume on the radiation loss by nitrogen and neon impurity seeding is studied in the SOLPS4.3 2D transport code. The study is carried out in a simplified tokamak geometry with long divertor legs, aimed at amplifying the effect of impurity radiation. It is shown that despite the simple analytical models predicting the enhancement of the low-Z impurity radiation loss with increasing cross-field heat conduction inside the divertor volume, the detailed 2D simulations demonstrate such a trend only if the electron density at the separatrix is used as the principal parameter quantifying the density of the edge plasma. However, if the total amount of the hydrogen isotope particles, which is a more suitable parameter for the edge plasma characterization, is considered, then the cross-field transport shows practically no impact on the impurity radiation in the divertor. It is found also that for a fixed ratio of the impurity to the main plasma content in the edge, two stable solutions can exist in a rather wide interval of the densities of the edge plasma. These solutions are attributed to the different distributions of the seeded impurity between the outer and inner divertors. Jumps between these two stable branches may result in bifurcation-like transitions between the attached and detached divertor regimes.