Cold Divertor Research Articles

SOLPS-ITER simulations of an X-point radiator in the ASDEX Upgrade tokamak

The X-point radiator (XPR) is an attractive scenario that may contribute to solving the power exhaust problem in future fusion devices. The 2D transport code SOLPS-ITER was applied to reproduce the experimentally measured plasma condition with an XPR in the ASDEX Upgrade tokamak and to compare with a reduced model. Neutrals penetrating from the adjoining cold divertor region and the large connection length near the X-point play an important role in initiating an XPR. However, once such a radiator is created, it persists even if the fueling and impurity seeding rates were reduced. The redistribution of plasma density and radiation near the X-point caused by fluid drifts at the XPR was studied in the simulation.

Structure, morphology and deuterium retention and release properties of pure and mixed Be and W layers

Deuterium retention and release behavior were investigated in this study, for beryllium, tungsten and mixed beryllium/tungsten thin layers which are of concern for next-generation thermonuclear fusion devices like International Thermonuclear Experimental Reactor. The layers prepared with thermionic vacuum arc technology by using two different ion acceleration voltages 0 V and −700 V respectively were subsequently implanted by exposure to a steady-state deuterium plasma with ion energies of 240 eV/D for D3 and 360 eV/D for D2 at a specific fluency of 2.85 × 1020 m−2 s. Morphology is strongly influenced by layer composition and by the ion acceleration voltage applied in-situ during layer deposition. Crystalline structure analysis shows the presence of a polycrystalline W metallic phase and also highlights a dependence between tungsten crystallite size and Be/W atomic ratio. The amount of deuterium in mixed layers is lower than in pure beryllium layers indicating that W is mitigating D binding states. Results extrapolated for a ‘cold chamber and divertor scenario’ indicate a limited desorption efficiency (∼20%), if the wall baking scenario, for tritium removal procedure in ITER, is to be applied to the investigated layers.

Thomson scattering measurements in the divertor region of the TCV Tokamak plasmas

In order to better study conventional and alternative divertor configurations, the TCV Tokamak was recently upgraded. In-vessel graphite baffles were installed in the TCV chamber to delimit a separate divertor chamber from the core that required enhanced diagnostic capabilities. To date, TCV's Thomson scattering diagnostic provided electron temperature and density profiles of the hot core and the pedestal of the plasma with an array of 96 polychromators. To study the new divertor region, an upgrade added 20 new polychromators specifically to diagnose the plasma in the relatively cold divertor region. New 4-channel polychromators were designed to provide Te measurements down to ∼1 eV with ne∼1.5× 1019 m−3. Interference filters of only 1.9 nm FWHM were required to achieve this goal. Te and ne measurements from the TCV divertor region have been obtained for the first time. Specific diverted plasma discharges where the outer divertor leg is scanned in and out of the TS laser beam's path are used to demonstrate the diagnostic's new capabilities and potential with measurements of Te as low as 1.4 eV obtained for electron densities of 2× 1018 m−3. The system's 16 mm integration length along the laser line is providing, for the first time, Te and ne profiles with sufficient spatial resolution across the divertor legs to resolve the divertor plasma dynamics.

Stabilization of EMC3-EIRENE for detachment conditions and comparison to SOLPS-ITER

Numerical access to cold divertor conditions in tokamaks with EMC3-EIRENE has been proven to be challenging, but is now possible for the first time after a linearization of the electron energy sink term from interaction with neutral gas. This is shown based on a 1D flux tube simulation, and consecutively for a full 3D ITER edge plasma configuration. Comparison with SOLPS-ITER simulations for the initial non-active phase in ITER with a gas fueled L-mode show good agreement with EMC3-EIRENE for the transition to detachment.

A study on the density shoulder formation in the SOL of H-mode plasmas

The term “shoulder formation” refers to an increase of the density decay length in the scrape-off layer (SOL) observed in many tokamaks during L-mode operation when a density threshold is reached. Recent experiments in ASDEX Upgrade (AUG) and JET have shown that the shoulder forms when the divertor collisionality in the divertor electrically disconnects filaments from the wall. This leads to a transition from the sheath limited to the inertial regime and to an enhancement of radial particle transport, in good agreement with analytical models. In the present work, the validity of such a mechanism is investigated in the more reactor-relevant H-mode regime. For this, a cold divertor H-mode scenario is developed in AUG using different levels of D puffing and N seeding, in which inter-ELM filaments and SOL density profiles are measured. The basic relation between filament size and divertor collisionality is still valid in H-mode plasmas, albeit an additional condition related to the gas fueling rate has been found for the formation of the shoulder.

Multi-fluid transport code modeling of time-dependent recycling in ELMy H-mode

Simulations of a high-confinement-mode (H-mode) tokamak discharge with infrequent giant type-I ELMs are performed by the multi-fluid, multi-species, two-dimensional transport code UEDGE-MB, which incorporates the Macro-Blob approach for intermittent non-diffusive transport due to filamentary coherent structures observed during the Edge Localized Modes (ELMs) and simple time-dependent multi-parametric models for cross-field plasma transport coefficients and working gas inventory in material surfaces. Temporal evolutions of pedestal plasma profiles, divertor recycling, and wall inventory in a sequence of ELMs are studied and compared to the experimental time-dependent data. Short- and long-time-scale variations of the pedestal and divertor plasmas where the ELM is described as a sequence of macro-blobs are discussed. It is shown that the ELM recovery includes the phase of relatively dense and cold post-ELM divertor plasma evolving on a several ms scale, which is set by the transport properties of H-mode barrier. The global gas balance in the discharge is also analyzed. The calculated rates of working gas deposition during each ELM and wall outgassing between ELMs are compared to the ELM particle losses from the pedestal and neutral-beam-injection fueling rate, correspondingly. A sensitivity study of the pedestal and divertor plasmas to model assumptions for gas deposition and release on material surfaces is presented. The performed simulations show that the dynamics of pedestal particle inventory is dominated by the transient intense gas deposition into the wall during each ELM followed by continuous gas release between ELMs at roughly a constant rate.

Investigation of conventional and Super-X divertor configurations of MAST Upgrade using scrape-off layer plasma simulation

One of the first studies of MAST Upgrade divertor configurations with scrape-off layer plasma simulation (SOLPS) 5.0 are presented. We focus on understanding the main prospects associated with the novel geometry of the Super-X divertor (SXD). This includes a discussion of the effect of magnetic flux expansion and volumetric power losses on the reduction of target power loads, the effect of divertor geometry on the divertor closure and the distribution of neutral species and radiation in the divertor, and the role of the connection length in broadening the target wetted area. A comparison in conditions typical for MAST inter-ELM H-mode plasmas confirms the improved performance of the Super-X topology resulting in significantly better divertor closure with respect to neutrals (the atomic flux from the target increased by a factor of 6, but the atomic flux from the divertor to the upper SOL reduced by a factor of 2), increased radiation volume and increased total power loss (a factor of 2) and a reduction of target power loads through both magnetic flux expansion and larger volumetric power loss in the divertor (a factor of 5–10 in attached plasmas). The reduction of the target power load by SXD further increases with collisionality (high density or detached regimes) thanks to the larger importance of volumetric power losses. It is found that a cold divertor plasma leads to stronger parallel temperature gradients in the SOL which drive more parallel heat flux, meaning that the effectiveness of perpendicular transport in spreading the power at the target can be reduced, and this needs to be taken into account in any optimization.

Evaluation of heat and particle controllability on the JT-60SA divertor

The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15MW/m2. Dependence of the heat flux mitigation on a D2 gas-puff is evaluated by SONIC simulations for high density (ne_ave∼1×1020m−3) high current plasmas. It is found that the peak heat load 10MW/m2 with dense (ned>4×1020m−3) and cold (Ted, Tid⩽1eV) divertor plasmas are obtained at a moderate gas-puff of Γpuff=15×1021s−1. Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100m3/s. In full non-inductive current drive plasmas with low density (ne_ave∼5×1019m−3), the reduction of divertor heat load is achieved with the Ar injection.

Quantification of chemical erosion in the DIII-D divertor and implications for ITER

The Porous Plug Injector (PPI) has proven to be an invaluable diagnostic for in situ characterization and quantification of erosion phenomena in DIII-D. Previous work has led to derivation of three primary figures of merit for chemical erosion (CE) in attached and cold divertor conditions: relative intensity of C+ chemical and physical sources, the CE yield (Ychem), and effective photon efficiencies for chemically eroded products. Application of these figures for accounting of observed absolutely calibrated CI and CII emission intensities is demonstrated to produce a self-consistent solution at the DIII-D targets. Reinterpretation of the CI (C0) spectral lineshape profile supports the relative roles of local chemical versus physical sputtering as previously determined for CII (C+). Comparison of calculated in situ Ychem to that measured ex situ suggests a tokamak-specific lower energy threshold for CE and has potentially major implications for prediction of tritium co-deposition near the divertor targets in ITER.

Characterization of chemical sputtering using the Mark II DiMES porous plug injector in attached and semi-detached divertor plasmas of DIII-D

An improved, self-contained gas injection system for the divertor material evaluation system (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH 4 are found to be 58 ± 12 in an attached divertor ( n e ∼ 1.5 × 10 13/cm 3, T e ∼ 25 eV, T surf ∼ 450 K), and 94 ± 20 in a semi-detached cold divertor ( n e ∼ 6.0 × 10 13/cm 3, T e ∼ 2–3 eV, T surf ∼ 350 K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C + at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH 4 from the PPI, then compared with that seen in intrinsic sputtering.

Modeling of dynamic response of SOL-divertor plasmas to an ELM crash

A dynamic five-point model is developed to study responses of scrape-off-layer (SOL) and divertor plasmas to a crash due to edge-localized modes (ELMs). The five-point model is beneficial to wide-ranging studies of SOL-divertor physics and to coupling with the core transport code in the integrated modeling. The five-point model can reproduce static and dynamic features obtained by usual fluid codes. The dynamic behavior obtained by the five-point model agrees fairly well with that by the particle code PARASOL. The influence of the ELM crash on the thermoelectric instability and the resultant asymmetry is investigated. The ELM crash transiently induces the thermoelectric instability and large SOL currents for asymmetric divertor plasmas before the ELM. The resultant currents drive convective heat flows and enhance the asymmetry on the fast time scale of electrons. The ELM crash is found to reverse the asymmetry before and after the ELM. Even for symmetric plasmas before the ELM, the asymmetry is induced on the slow time scale of ions by an ELM crash, because the transient formation of dense and cold divertor plasmas causes the thermoelectric instability.

Plasma-Surface Interaction Studies on DIII-D and Their Implications for Next-Step Fusion Experiments

Unique diagnostic and access features of the DIII-D tokamak, including a sample exposure system, have been used to carry out controlled and well-diagnosed plasma-surface interactions (PSI) experiments. An important contribution of the experiments has been the ability to link a given plasma exposure condition to a measured response of the plasma-facing surface and to thus understand the interaction. This has allowed for benchmarking certain aspects of erosion models, particularly near-surface particle transport. DIII-D has empirically quantified some of the PSI effects that will limit the operation availability and lifetime of future fusion devices, namely, net erosion limiting divertor plate lifetime and hydrogenic fuel retention in deposit layers. Cold divertor plasmas obtained with detachment can suppress net carbon divertor erosion, but many low-temperature divertor PSI phenomena remain poorly understood: nondivertor erosion sources, long-range particle transport, global erosion/deposition patterns, the enhancement of carbon erosion with neon impurity seeding, the sputtered carbon velocity distribution, and the apparent suppression of carbon chemical erosion in detachment. Long-term particle and energy fluences have reduced the chemical erosion yield of lower-divertor tiles. Plasma-caused modification of a material’s erosion properties, including material mixing, will occur quickly and be important in long-pulse fusion devices, making prediction of PSI difficult in future devices.

Reduction of divertor heat load in JET ELMy H-modes using impurity seeding techniques

The main objective of this paper is investigation of methods for reduction of divertor heat loads in order to increase the lifetime of divertor tiles in future fusion reactors. Special emphasis is given to studies of reduction of transient heat loads due to edge localized modes (ELMs). Two methods are compared: argon seeded type-I ELMy H-modes and nitrogen seeded type-III ELMy H-modes. In both scenarios, the impurity seeding leads to a reduction in the pedestal energy and hence a reduction in the energy released by the ELM. This consequentially reduces the power load to the divertor targets. At high radiative power fractions in type-III ELMy H-modes, part of that released ELM energy (small ELMs, below 20 kJ) is dissipated by radiation in the scrape off layer (SOL). Modelling of the ELM mitigation supports the experimental findings. This ELM mitigation by radiative dissipation is not effective for larger ELMs. In between ELMs, the plasma is detached and radiates strongly from the X-point region. During an ELM, the nitrogen in the X-point and divertor region becomes ionized into more weakly radiating higher charge states and the plasma re-attaches for large ELMs. At JET, argon radiates predominantly in the main plasma and not so much in the cold divertor region. Hence, the effect of radiative dissipation of ELM heat fluxes by argon is very low due to the limited argon density in the divertor region. Nevertheless, both scenarios might be compatible with an integrated ITER scenario, with respect to acceptable divertor lifetime and acceptable confinement.

Characteristics of boundary plasmas after the removal of divertor baffle plates and boronization on JFT-2M

A dense and cold divertor ( n e div∼2×10 19 m −3, T e div∼6 eV) during H-mode, was realized with the closed divertor (CD2) on JFT-2M. By removing all baffle plates, such a condition was compared between open (OD) and closed configurations. When a strong gas-puff is introduced into the divertor region during L-mode, the baffling effect of CD2 is verified definitely by the reduction of neutral particle backflow and the enhanced radiation localization in the divertor region. Introducing a strong gas-puff during H-mode for the OD case, H-mode is terminated in an early stage when the radiation loss fraction exceeds a critical level. A dense and cold divertor, as shown in the CD2 case, is not obtained. This is attributed to the lack of particle-handling capability for the OD compared with the CD2 configuration. As an initial result, the radiation loss fraction was reduced by half after the boronization which prevents the H–L transition.

Simulation of divertor detachment characteristics in JT-60 with superconducting coils

The capability of detachment control in JT-60SC is demonstrated with a divertor code (SOLDOR/NEUT2D). Under the standard operation of the core edge density at 95% of the minor radius, n edge=3.2×10 19 m −3, the power flow from the core plasma Q e= Q i=6 MW, and the conductance of inner and outer cryopanels C in= C out=50 m 3/s, the inner divertor plasma is partially detached and the outer divertor plasma attached. The inner divertor can be changed from partially detached to attached plasma by inner pump of 150 m 3/s. The fully detached inner divertor is obtained in high density operation with n sep>4×10 19 m −3, where n sep is the density at the separatrix outer midplane, The elastic collision in a dense and cold divertor plasma ( n e⩾5×10 20 m −3, T e<5 eV) play an important role on detachment.

Deuterium pumping experiment with superpermeable Nb membrane in JFT-2M tokamak

A new divertor pumping system with superpermeable membranes has been installed in the JFT-2M tokamak to investigate the applicability of membrane pumping to fusion devices. The membrane pumping is based on suprathermal (>1 eV) hydrogen permeation through a niobium membrane. Deuterium pumping is observed with divertor plasmas in tokamak environment. The pumping flux through the membrane increases by NBI heating and additional gas puffing and it is found that the pumping flux is directly proportional to the pressure in the divertor chamber where the pumping system is placed. The maximum permeation flux density is 7.3×10 19 D/m 2 s for a tokamak discharge with a dense and cold divertor plasma.

Study of a closed divertor with strong gas puffing on JFT-2M

In order to improve the particle control capability of the divertorand to demonstrate the possibility of a dense and cold divertor withhigh confinement plasmas, a closed divertor study has been carried outon the JFT-2M tokamak. When a strong gas puff is applied into thedivertor chamber, a baffling effect is revealed by enhanced radiationlocalization and high neutral pressure in the divertor chamber, lowcore fuelling and sustainment of core confinement quality. Thebaffling effect is further enhanced with E × B flow and/or current inthe scrape-off layer by applying divertor biasing. A dense and colddivertor state (nediv ≈ 4 × 1019m-3 and Tediv ≈ 4 eV), together with the improved confinement modes, can be obtained by strong gas puffing. Furthermore, the improved core confinement is not affectedsignificantly in the high density region (n̄e/neG < 0.7). TheUEDA code simulation reproduces the baffling effect and a dense andcold divertor plasma with a closed divertor structure.

Carbon influx in He and D plasmas in DIII-D

Differences in the carbon behavior between He and D plasmas during VH-mode, L-mode and L-mode with excess gas puffing are reported and inferences on the importance of the various carbon sources during these modes of operation are discussed. During a VH-mode phase, VUV and visible charge exchange spectroscopy indicates that for both He and D operation the carbon behavior is very similar. In the edge plasma, carbon build up is quite rapid, and the carbon influx represents a large fraction of the total plasma density increase until the termination of the VH phase. During cold divertor operation induced by puffing the primary fueling gas, D and He discharges show a difference in the carbon behavior. The core carbon density is seen to be approximately constant during a D discharge as it transitions from an attached to a cold divertor. However in a He discharge, the core carbon density disappears soon after the cold divertor transition. Arguments are made that the primary carbon source in the ELM free H-mode period is physical sputtering by ion impact at the divertor strike point. In L-mode, both attached and cold divertor, the primary source is from the divertor region and two possibilities for this source are chemical sputtering or charge exchange neutral sputtering. Existing data supports charge exchange neutrals as dominant.

Compatibility of H-mode with a radiating boundary and divertor detachment

ITER or any other future fusion device will operate in H-mode to take advantage of the improved confinement. Stationary operation in H-mode, however, requires the occurrence of ELMs, and the resulting power load on the divertor target plates is not acceptable from the engineering point of view. Beyond this, the present divertor concepts rely on a cold and dense (possibly detached) divertor plasma. Large type-I ELMs, however, carry too large a power and possibly burn through such cold divertor plasmas. The completely detached H-mode (CDH-mode), recently established in ASDEX Upgrade shows detachment at both divertor target plates and small ELMs (so-called type-III ELMs), that do not burn through the divertor plasma. The extrapolation of this scenario to ITER, however, is an important question.

Characteristics of and issues regarding combined H-modes

Owing to the recent progress of `profile controls' utilizing the freedom of profiles of plasma parameters, various types of combined (advanced) H-modes have been obtained; PEP H-mode, high- H-mode, high- H-mode, high- H-mode, reversed shear H-mode, CH-mode, VH-mode, etc. These combined H-modes with high potential for confinement and stability are characterized by improved transport in the core region in addition to the edge or deep inward penetration of the edge confinement pedestal. Such additional improvements seem to be related to flow shear, magnetic shear, safety factor, density profile or . The improved confinement region can propagate from the core to the edge or from the edge to the core, or appear almost simultaneously over the whole radius. However, there are some critical issues to be solved concerning reactor conditions, that is, under electron heating at high edge density with small impurity accumulation in the steady-state. Most of the NB heated combined H-modes have been obtained with ion heating, where ion thermal diffusivity and particle diffusivity can be reduced to the neoclassical level. However electron thermal diffusivity has so far not been clearly reduced except in the H-mode and in the VH-mode. Therefore, improvement of electron transport is one of the main issues. The second issue is that some improved modes are accompanied by strong density peaking which may result in large impurity accumulation. The third issue is the difficulty in achieving the improved modes at a high edge density with the high particle recycling essential for a dense and cold divertor. The fourth issue is the stability at high in the steady state. It is not certain that pressure and current profiles including the bootstrap current can be sustained stably. The requirement of ELMs for heat and particle exhaust is also an important issue in stability design as concerns the extent of the second regime access for the high-n ballooning mode.