Search

In attaining the cold plasma condition at the divertor target via gas puffing and/or nitrogen seeding, a large variation in the pressure-momentum and cooling losses has, for the first time, been demonstrated on JET with the ITER-like wall, leveraging novel target electron temperature measurements and JET’s unique capability for disentangling the main-ion-driven net dissipation effects from impurity radiation. These observations provide key insights into the dependence of volumetric losses on the dominant dissipation mechanisms, whether due to neutral-plasma interaction processes, impurity radiation induced detachment, and their interplay.While numerical code studies indicate that both volumetric cooling and pressure-momentum losses in the scrape-off layer (SOL) are essential for detachment to occur, experimental confirmation has been largely lacking, in large part due to diagnostic limitations. In the present assessment of L-mode discharges in the diagnostically optimized outer horizontal target configuration, the onset of pressure-momentum losses in nitrogen seeding scans in low recycling conditions is shown to occur at relatively high target electron temperature of 10 eV compared to pure deuterium pressure scans, leading to a more pronounced reduction in the target heat flux, and demonstrating for the first time the importance of momentum losses in impurity radiation dominated dissipative divertors in the absence of strong neutral-plasma interactions. Conversely, a relatively constant target heat flux response is observed above Te,t = 5 eV in pure deuterium density ramps, where the decrease in Te,t is balanced by a rise in target pressure prior to the ion flux rollover, coinciding with the pressure-momentum loss onset. While in practice the relative strength of each dissipation channel will be driven by integrated scenario requirements and core plasma performance constraints, this work demonstrates that there is no universal form of the fmom-loss and fcooling volumetric loss factors and their correlation. Moreover, the results confirm the theoretical and code predictions that both volumetric loss processes are required for detachment, while also resolving an apparent, but counter-intuitive, finding of the code results that volumetric power loss is insensitive to the presence of impurities, demonstrating that this is not the case experimentally. Further interpretive modelling is required to elucidate the dominant momentum and cooling loss channels in the observed trends.

Investigations of parametric scaling of power exhaust in the alternative divertor configuration (ADC) SOLPS-ITER simulation database of the EU-DEMO are conducted and compared to predictions based on the Lengyel model. The Lengyel model overpredicts the necessary argon concentrations for LFS divertor detachment by about a factor of 5–10 relative to the SOLPS-ITER simulations. Therefore, while the Lengyel model predicts that plasmas with acceptable divertor heat loads in EU-DEMO would exceed the tolerable upstream impurity concentrations by a large margin, there are several SOLPS-ITER solutions within an acceptable operational space. The SOLPS-ITER simulations indicate that, unlike assumed by the standard Lengyel model, there are significant heat dissipation mechanisms other than argon radiation, such as cross-field transport, that reduce the role of argon radiation by a factor of 2 to 3. Furthermore, the Lengyel model assumes that the radiation front is powered by parallel heat conduction only, which tends to lead to a narrow radiation front as the radiative efficiency increases strongly with reducing thermal conductivity. As a result, the radiative volume and total impurity radiation are suppressed for a given impurity concentration. However, the SOLPS-ITER simulations indicate that other mechanisms, such as cross-field transport, can compete with parallel heat conduction within the radiative front and increase the radiative volume. As a result, the standard Lengyel model provides a very pessimistic estimate for the necessary concentration for impurities with strong radiative capability in low temperatures around 10 eV, such as argon. However, parametric scaling relations are needed for fast scoping of the operational space without having to run a complex, numerical code, such as SOLPS-ITER. In these applications, the Lengyel model might still be useful to provide guidance on the relative scaling of the exhaust characteristic of the various points in the operational space. The aim of this work is to provide evidence to avoid overconfidence on these simple model predictions.

A double-null configuration is being considered for the EU-DEMO, due to its potential benefits for power exhaust arising from the use of two active divertors and magnetically disconnected low- and high-field sides. Using systematic parameter scans in fluid simulations, we have investigated the divertor power exhaust in the EU-DEMO in a connected double-null configuration, and compared the edge plasma properties to those obtained in a single-null configuration under detached conditions anticipated for reactor operation. Neglecting drift effects and kinetic behaviour of the neutrals, no clear benefits of the double-null configuration could yet be identified for the radiation pattern and power mitigation on open field lines. Future work should address the aforementioned physics as well as the effect of the additional X-point on core radiation.

A physics and engineering analysis of alternative divertor configurations is carried out by examining benefits and problems by comparing the baseline single null solution with a Snowflake, an X- and a Super-X divertor. It is observed that alternative configurations can provide margin and resilience against large power fluctuations, but their engineering has intrinsic difficulties, especially in the balance between structural solidity and accessibility of the components and when the specific poloidal field coil positioning poses further constraints. A hybrid between the X- and Super-X divertor is proposed as a possible solution to the integration challenge.

The standard Single Null (SN) divertor is currently expected to be installed in DEMO. However, a number of alternative configurations are being evaluated in parallel as backup solutions, in case the standard divertor does not extrapolate successfully from ITER to a fusion power plant. We used the SOLPS code to produce a preliminary analysis of two such configurations, the X-Divertor (XD) and the Super X-Divertor (SX), and compare them to the SN solution. Considering the nominal power flowing into the SOL (PSOL = 150 MW), we estimated the amplitude of the acceptable DEMO operational space. The acceptability criterion was chosen as plasma temperature at the target lower than 5eV, providing low sputtering and at least partial detachment, while the operational space was defined in terms of the electron density at the outboard mid-plane separatrix and of the seeded impurity (Ar only in the present study) concentration. It was found that both the XD and the SXD extend the DEMO operational space, although the advantages detected so far are not dramatic. The most promising configuration seems to be the XD, which can produce acceptable target temperatures at moderate outboard mid-plane electron density (nomp=4.5×1019 m−3) and Zeff= 1.3. Keywords: DEMO, Advanced divertor, Numerical modeling, Detachment

Robust power exhaust schemes employing impurity seeding are needed for target operational scenarios in present day tokamak devices with metallic plasma-facing components (PFCs). For an electricity-producing fusion power plant at power density Psep/R>15MW/m divertor detachment is a requirement for heat load mitigation. 2D plasma edge transport codes like the SOLPS code as well as plasma-wall interaction (PWI) codes are key to disentangle relevant physical processes in power and particle exhaust. With increased quantitative credibility in such codes more realistic and physically sound estimates of the life-time expectations and performance of metallic PFCs can be accomplished for divertor conditions relevant for ITER and DEMO. An overview is given on the recent progress of plasma edge and PWI modelling activities for (carbon-free) metallic devices, that include results from JET with the ITER-like wall, ASDEX Upgrade and Alcator C-mod. It is observed that metallic devices offer an opportunity to progress the understanding of underlying plasma physics processes in the edge. The validation of models can be substantially improved by eliminating carbon from the experiment as well as from the numerical system with reduced degrees of freedom as no chemical sputtering from amorphous carbon layers and no carbon or hydro-carbon transport are present. With the absence of carbon as the primary plasma impurity and given the fact that the physics of the PWI at metallic walls is less complex it is possible to isolate the crucial plasma physics processes relevant for particle and power exhaust. For a reliable 2D dissipative plasma exhaust model these are: cross-field drifts, complete kinetic neutral physics, geometry effects (including main-chamber, divertor and sub-divertor structures), SOL transport reflecting also the non-diffusive nature of anomalous transport, as well as transport within the pedestal region in case of significant edge impurity radiation affecting pedestal pressure and hence Psep.

We have studied the retention of nitrogen in surfaces with varying roughness and the transport of nitrogen from the divertor to the outer midplane in experiments at the tokamak ASDEX Upgrade. To allow for a reliable identification of nitrogen retained from the plasma exposure, 5.3 · 1021 atoms of the tracer isotope 15N were injected into the private flux region of the plasma. On polished W samples exposed to the plasma in the outer divertor, the N content peaks to both sides of the strike line with an areal density of 1.5 · 1020 15N/m2 and drops to a value of 1.0 · 1020 15N/m2 in the strike line region. In contrast, the N content of samples with a rougher surface peaks at the strike line and reaches areal densities of 3.0 · 1020 15N/m2. The N deposition at the outer limiters was measured via samples exposed on the midplane manipulator. At the limiter position the 15N areal density reaches a value of 0.2 · 1020 15N/m2, only a factor of ten smaller than the areal densities in the divertor. However, a comparison to WallDYN simulations shows that the observed N content is roughly a factor of 4 smaller than the value predicted by these simulations. Possible reasons for this discrepancy are the toroidal asymmetric main wall geometry, which currently cannot be fully included in the simulations, or an enhanced re-erosion of deposited N. Keywords: Tokamak, Impurity migration, Nitrogen, Roughness, WallDYN