Divertor Channel Research Articles

Study of passively stable, fully detached divertor plasma regimes attained in innovative long-legged divertor configurations

Numerical modeling of divertor configurations with radially or vertically extended, tightly baffled, outer divertor legs has demonstrated the existence of a passively-stable fully detached divertor regime. In the simulations, long-legged divertors provide up to an order-of-magnitude increase in peak power handling capability compared to conventional divertors. The key physics for attaining the passively stable, fully detached regime in these simulations involves the interplay of strong convective plasma transport to the divertor leg outer sidewall, confinement of neutral gas in the divertor volume, geometric effects including a secondary X-point, and atomic radiation. New analysis shows that in this regime the detachment front location is set by the balance between the power entering the divertor leg and the losses to the walls of the divertor channel. Correspondingly, the maximum power that can be accommodated by the divertor, while still staying detached, increases with the poloidal length of the leg. The detached regime access window in terms of input power, density and impurity seeding concentration varies quantitatively depending on divertor geometry and modeling assumptions most specifically, cross-field transport to the side walls.

Conduction-based model of the Scrape-Off Layer power sharing between inner and outer divertor in diverted low-density tokamak plasmas

Abstract A simple analytic model for the repartition of the Scrape-Off Layer (SOL) exhaust power between the inner and outer divertors in a diverted low-density tokamak plasma is introduced. Electron heat conduction is assumed to dominate the heat transport, from the outboard mid-plane to the divertor targets, with no heat sinks or sources in the SOL. Both divertor channels are in the attached, high-recycling regime. The model is in reasonable qualitative agreement with recent TCV experimental data and EMC3-Eirene simulations. For the Single Null divertor, it reproduces the experimentally observed increase in the power ratio between the inner and outer divertor plates of TCV, P in / P out , with increasing the outer divertor leg length or the outer target flux expansion. For the Snowflake Minus configuration, it reproduces the observed variation of P in / P out with X-point separation, although only for the reversed magnetic field direction. Within the model limitations, it provides a basic understanding of the power sharing in alternative divertor geometries.

UEDGE modelling of detached divertor operation for long‐leg divertor geometries in ARC

Simulations of the ARC scrape‐off layer and divertor configuration are carried out using UEDGE to perform the first power‐handling assessment for the divertor design. A convective–diffusive anomalous transport model is implemented in the code to reproduce the radial plasma profiles anticipated for the device at the outer mid‐plane. Levels of exhaust power from the core are varied, and a 0.5% fixed fraction neon impurity seeding is included in investigations to find steady‐state detached solutions. Initial studies employing a super‐X divertor configuration have shown that a stable detached divertor‐operational window exists for reactor exhaust powers in the range of 72–92 MW, which is close to projected levels of exhaust power of 105 MW across the separatrix anticipated for ARC. Under these conditions, the divertor heat flux is fully dissipated by radiation and radial losses to the sidewalls of the divertor channel, leading to acceptable levels of power flux density on plasma‐facing components. Simulations are being extended to study the performance of the X‐point target geometry and to explore the sensitivity to the physics model and input parameters.

The snowflake divertor

The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. Among potential beneficial effects of this geometry are: increased volume of a low poloidal field around the null, increased connection length, and the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described.

An Energy Recovering Divertor Based on Amplification of Alfven Waves

A new technique for direct power extraction from a fusion plasma is presented. This technique is designed to operate within a tokamak divertor channel without requiring plasma to cross the toroidal field coils, and is therefore termed an energy recovering divertor. The proposed technique is based on amplification of Alfven waves by ion kinetic energy, using a mechanism analogous to a free-electron laser. Presented here are a detailed description of the technique and analytic calculations of its basic mode of operation.

Overview of material re-deposition and fuel retention studies at JET with the Gas Box divertor

In the period 1998—2001 the JET tokamak was operated with the MkII Gas Box divertor. On two occasions during that period a number of limiter and divertor tiles were retrieved from the torus and then examined ex situ with surface sensitive techniques. Erosion and deposition patterns were determined in order to assess the material erosion, material migration and fuel inventory on plasma facing components. Tracer techniques, e.g. injection of 13C labelled methane and tiles coated with a low-Z and high-Z marker layer, were used to enhance the volume of information on the material transport. The results show significant asymmetry in the distribution of fuel and plasma impurity species between the inner (net deposition area) and the outer (net erosion) divertor channels. No significant formation of highly hydrogenated carbon films has been found in the Gas Box structure. The important processes for material migration, and the influence of operation scenarios on the morphology of the deposits are discussed. Comparison is also made with results obtained following previous divertor campaigns.

Effect of graphite and molybdenum wall tile reflections on visible light diagnostics in tokamak experiments

Visible light reflectivities are measured for graphite and molybdenum tokamak wall tiles as a function of incident angle and wavelength. Using the measured reflectivities together with a ray-tracing routine, estimates of the effect of wall reflections on Dα light measurements in the DIII-D and Alcator C-Mod tokamaks are made. It is found that the contribution of reflections is small (typically <10%) for divertor view channels, but can be significant (>50%) for other view chords. Of the nondivertor channels, near-midplane view chords are found to be least susceptible to reflections from the divertor and are therefore best suited for main chamber recycling measurements.

Safety implications of an integrated boiling curve model for water-cooled divertor channels : Marshall, T. Fusion Engineering and Design, 2002, 63–64, 235–242

Safety implications of an integrated boiling curve model for water-cooled divertor channels : Marshall, T. Fusion Engineering and Design, 2002, 63–64, 235–242

Lithium surface operating under steady-state power load

A liquid lithium surface is considered for application in divertor of a fusion tokamak-reactor. Lithium surface has been realized in experimental mock-ups and its operation has been demonstrated under high power load at reactor relevant heat fluxes. Lithium targets have been developed on the basis of capillary pore structures. A vertical working surface was investigated under steady-state electron beam. The range of power loads 1–50 MW/m 2 was covered by the studies. Long-duration experiments were performed on thermally stabilized targets at 1–10 MW/m 2. Evaporation was shown to be efficient mechanism of power removal and a high lithium mass loss rate was measured. Operation of the facility with plasma at 0.2 g/s of lithium flow was shown. The problem of lithium balance in divertor and SOL is discussed. Pumping of lithium is possible by solid and liquid metal wall structures in reactor conditions in the divertor channel.

Safety implications of an integrated boiling curve model for water-cooled divertor channels

The international fusion community is actively researching advanced heat transfer methods for removal of high thermal loads from next-generation divertor assemblies. Such advanced techniques may indeed optimize the operational and economical performance of future divertor designs. However, with its extensive operational database, water-cooling remains as one of the optimum choices for near-term divertor designs. Critical heat flux (CHF) is the maximum heat flux that water, at a given set of inlet conditions, can remove via fully developed nucleate boiling. Accordingly, an accurate CHF calculation is of the utmost importance for maintaining adequate safety margins in divertor operation. This paper uses the integrated boiling curve model developed at Sandia National Laboratories to examine the safety implications of calculating the CHF. In particular, this paper focuses on the influence of the finite element peaking factor (FEPF) that converts the heat flux predicted by CHF correlations into a plasma heat flux that can be measured. The analyses illustrate that the FEPF is proportional to the plasma heat flux and thus accurate calculation of the CHF requires the use of the appropriate FEPF for the given water conditions and plasma heat flux. It is shown that using a geometric peaking factor is inadequate since the true peak factor is dependent upon the plasma heat flux. The conclusion is that a finite element analysis incorporating an integrated boiling curve is required for accurate calculation of the CHF.

Guidance of Divertor Channel by Cusp-Like Magnetic Field for Tokamak Devices

A divertor magnetic configuration is proposed that significantly reduces heat load on the divertor plates in tokamak devices. The proposed configuration utilizes an additional cusp-like magnetic fi...

Modeling the Nukiyama Curve for Water-Cooled Fusion Divertor Channels

A conclusive safety assessment of a fusion reactor requires that the thermal response of the divertor assembly is known with a high degree of accuracy. Such accuracy is mandated because the divertor assembly is subjected to the highest levels of incident heat flux within the reactor. In order to accurately predict the thermal response of the divertor's cooling channels, it is necessary to have a complete model of the Nukiyama boiling curve for the water conditions of interest. Currently published models of the Nukiyama curve for fusion divertor channels have only included the regimes of forced convection, partially and fully developed nucleate boiling, and the local CHF. This paper presents a model that includes these pre-CHF regimes and the post-CHF regime of transition boiling. The model is unique because (1) it tightly integrates the respective heat transfer correlations and makes heat transfer predictions for the water conditions and incident heat fluxes that are fusion-specific, (2) predicts post-CHF heat transfer properties for a swirl tape divertor channel, and (3) validates its predictions via comparison with experimental data. Based on these three points, this model is considered as one of the best available methods for predicting the Nukiyama curve for a water-cooled fusion device.

One-dimensional simulation on stability of detached plasma in a tokamak divertor

The stability of the radiation front in the scrape-off layer (SOL) of a tokamak is studied with a one-dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. As the energy from the core into the SOL plasma is reduced, stable attached solutions change into stable detached solutions. Whenever such stable detached states are attained, the strong radiation front is in contact with or at a small distance from the divertor target. When the input energy is decreased further and total losses become larger than the input power, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such a case, no solutions for the radiation front resting in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of the radiation front.

An Experimental Examination of the Loss-of-Flow Accident Phenomenon for Prototypical ITER Divertor Channels of Y = 0 and Y = 2

This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups’ survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pump was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.

Operational performance of JT-60U W-shaped divertor

Performance of inner leg pumping of JT-60U W-shaped divertor is presented. Inner leg pumping works well as the dome separates inner and outer divertor channels for neutral particles in the private flux region. In ELMy H-mode with NBI heating power of 18 MW, the operational regime where inner leg pumping is effective extends to the divertor radiation range of about 60% of SOL power. The fraction of inner particle flux pumped is estimated to be about 3%. It is qualitatively shown that a combination of main gas puff and inner leg pumping reduces carbon impurities in the main plasma, suggesting the existence of SOL flow effect on impurity reduction.

Momentum and Energy Transfer Via Neutral Atoms and Molecules in an International Thermonuclear Experimental Reactor Low-Pressure (10-mTorr) Gas Target Divertor

One option for particle and power handling in the International Thermonuclear Experimental Reactor (ITER) is the creation of a low-pressure (∼10-mTorr) gaseous divertor. The divertor would have a long channel over which energy would be removed from the plasma by radiation, and the plasma pressure would be balanced by a change inflow velocities and neutral pressures entering the sides of the channel This combination should substantially reduce the ion energy and ion flux that impact the eventual end of the divertor channel. For this concept to work, momentum must be removed from the plasma by the neutral atoms and molecules. Plasma parameters were taken from a DDC83 code solution. A Monte Carlo treatment of the plasma-neutral interactions has been obtained using DEGAS, which includes charge-exchange, recombination, ion-neutral, and neutral-neutral elastic collisions. Results show that the momentum transferred to the side walls is insufficient by two orders of magnitude to achieve the pressure reduction needed. Each molecule that enters the plasma makes hundreds of elastic and inelastic collisions in the plasma and then is more likely to be ionized (transferring the momentum back to the plasma) than to travel to a wall.

Simulation of Neutral Atoms and Molecules in an International Thermonuclear Experimental Reactor High-Pressure (1-Torr) Ultrahigh Recycling Divertor

The Monte Carlo code DEGAS was used to investigate the neutral atom and molecular interactions for a high-pressure (∼1-Torr) gaseous divertor in the International Thermonuclear Experimental Reactor (ITER). Energy is removed from the plasma by radiation while the plasma pressure is balanced predominantly by a high neutral pressure at the end of the divertor. Plasma parameters were taken from the two-dimensional fluid code PLANET. Neutral sources from both ions recycling off the walls and recombination were included. The neutral density peak calculated with DEGAS of 3.43 ± 0.01 × 1022 m−3 occurred 4.5 cm from the divertor channel end. The ion and neutral atom energy fluxes were calculated to determine the heat load onto the divertor walls. A code was written to calculate the radiation distribution onto the side walls, not including any radiative absorption or reemission. The total energy flux peak (including ions, neutrals, and radiation) was 4.28 ± 0.30 MW/m2. This falls below the design criteria of 5 MW/m2. These results may help determine the wall material, heat removal, and the vacuum pumping requirements for the ITER divertor design and show the importance of a full treatment of neutral atoms and molecules in these regimes.

Plasma-wall interactions in ITER

This paper reviews the status of the design of the divertor and first-wall/shield, the main in-vessel components for ITER. Under nominal ignited conditions, 300 MW of alpha power will be produced and must be removed from the divertor and first-wall. Additional power from auxiliary sources up to the level of 100 MW must also be removed in the case of driven burns. In the ignited case, about 100 MW will be radiated to the first wall as bremsstrahlung. Allowing the remaining power to be conducted to the divertor target plates would result in excessive heat fluxes. The power handling strategy is to radiate an additional 100–150 MW in the SOL and the divertor channel via a combination of radiation from hydrogen, and intrinsic and seeded impurities. Vertical targets have been adopted for the baseline divertor configuration. This geometry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densities on the target plates can be ≤ 5 MW/ m 2. Such regimes promote relatively high pressure (> 1 Pa) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping speed to pump helium is ≤ 50 m 3/ s. An important physics question is the quality of core confinement in these attractive divertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performance of in-vessel components. Both centered disruptions and VDE's produce stresses in the first-wall/shield modules, backplate and the divertor wings and cassettes that are near or even somewhat in excess of allowables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties are major factors determining component lifetime. Considering the potential for impurity contamination and minimizing tritium inventory as well as thermomechanical performance, the present material selection calls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux could be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the final selection for the first assembly will be made at the end of the EDA.

Simulation of the Alcator C‐Mod Divertor with an Improved Neutral Fluid Model

AbstractWe have simulated detachment in Alcator C‐Mod, using an improved neutral fluid model. It features a parallel neutral momentum equation, fully coupled to the ions, allowing parallel plasma momentum to be converted into neutral momentum, which is dispersed to the material walls due to the high neutral viscosity. The simulations reproduce the pervasive experimental feature that the plasma detaches gradually, starting at the strike point, while always remaining attached above the nose of the divertor channel. The strike point heat flux and current drop by an order of magnitude as 0.5% of carbon is introduced to induce partial detachment. The plate temperature remains at or above about 1 eV all the way out to the divertor nose, in agreement with the experimental data.

Progress in the design of in-vessel components for ITER

We review the design of in-vessel components for ITER, specifically the vacuum vessel, divertor, first wall and blanket-shield, and the r.f. coupling systems. The vacuum vessel is the most straightforward in concept but also the most critical since it forms the primary containment boundary. The design is a double-wall structure supported by radial plates and filled with steel balls which are directly water cooled. The structure is typically 40–70 cm thick and the shielding performance of the vessel together with the blanket-shield is sufficient to reduce the nuclear heating in the magnets to less than 8 kW. The vessel is designed to withstand an over-pressure of 2 MPa. The nominal fusion power is specified to be 1500 MW; the first wall and the divertor are each being designed to remove 80% of the corresponding a power. The conventional high recycling approach to the divertor leads to power densities two to four times higher than can be removed by practical designs for the divertor targets. However, many divertor experiments have shown that partial or full detachment of the divertor plasma from the target plates is possible in high density regimes. In such cases, the peak power density can be reduced by a factor of 5 or more, and the bulk of the power is then radiated in the divertor channel, near the X point or in the scrape-off layer. Thus the first wall and divertor are being designed to accommodate any of these possibilities. Edge localized modes and disruptions can burn through the protective plasma-gas cushion in front of the target plates, and erosion remains a major concern. While beryllium has been selected as the first-wall material, the selective use of tungsten in areas of intense charge exchange flux is under consideration. Earlier design efforts were concentrated on an integrated first wall and blanket-shield, but recent emphasis has been on an approach with a separately cooled first wall. Heating and current drive for ITER could be furnished by ion cyclotron r.f. (ICRF) or electron cyclotron r.f. (ECRF) heating, or by neutral-beam injection (NBI). All these options are being considered. ICRF is the option at present favored in view of its well-developed technology. However, the antennae become part of the first wall and their design must address reliability and maintenance issues as well as the effects of high thermal and disruption loads. ECRF and NBI are more attractive from this point of view, but their feasibility hinges on the development of suitable sources.