Closed Divertor Configuration Research Articles

Bayesian inference of particle source and sink in a closed-divertor using Balmer line spectroscopy

A new analysis technique for Balmer line spectroscopy that enables recombination rate (particle sink) and ionization rate (particle source) inference in a closed divertor configuration is reported. Bayesian inference is employed to systematically utilize all available information from multiple Balmer lines and constrain parameter ranges by using prior knowledge about plasmas. While a closed-divertor facilitates detachment, neutral plugging typically leads to large spatial variations in plasma parameters. A forward model is developed to take into account non-uniformity in the plasma parameters and applied to test data generated by divertor plasma simulations. It is shown that the forward model robustly provides particle source and sink inference over a wide parameter range. In addition, the precision improves as more Balmer lines are resolved simultaneously. The new analysis technique is also applied to an L-mode ASDEX Upgrade plasma in the high-recycling regime. The inferred quantities and their profiles are consistent with the expectations of a high-recycling divertor plasma. The further insight into the detachment physics will be provided by using this new analysis technique.

Density dependence of SOL power width in ASDEX upgrade L-Mode

Understanding the heat transport in the scrape-off layer (SOL) and the divertor region is essential for the design of large fusion devices such as ITER and DEMO. Current scalings for the power fall-off length λq in H-Mode [1] are available only for the outer divertor target at low densities with low recycling divertor conditions. For the divertor power spreading S only an empirical scaling for ASDEX Upgrade L-Mode is available based on global plasma parameters [2]. Modelling using SOLPS shows a dependence of S on the divertor electron temperature [3]. A more detailed analysis of the heat transport forming λq and S is presented for ASDEX Upgrade L-Mode discharges in hydrogen (H), deuterium (D) and helium (He). For low densities the power fall-off length λq,o on the outer divertor target in H and D is described by the same parametric dependencies as the H-Mode scaling [1] but with a larger absolute size of the power fall-off length in L-Mode.The divertor power spreading S is studied using the local divertor measurements of the target electron temperature Te,tar and density ne,tar. It is found that the competition of the diffusive transport parallel and perpendicular to the magnetic field forming S∝χ⊥/χ∥ is dominated by the temperature dependence of parallel electron conduction. For high divertor temperatures the ion gyro radius has a significant contribution to S, resulting in a minimum of S at ∼30 eV.A recent study [4] with an open divertor configuration found an asymmetry of the power fall-off length between inner and outer target with a smaller power fall-off length λq,i on the inner divertor target. Measurements with a closed divertor configuration find a similar asymmetry for low recycling divertor conditions. It is found, in the experiment, that the in/out asymmetry λq,i/λq,o is strongly increasing with increasing density. Most notably the heat flux density at the inner divertor target is reducing with increasing λq,i whilst the total power onto each divertor target stays constant. It is found that λq,o exhibits no significant density dependence for hydrogen and deuterium but increases with about the square root of the electron density for helium. The difference between H,D and He could be due to the different recycling behaviour in the divertor. These findings may help current modelling attempts to parametrize the density dependence of the widening of the power channel and thus allow for detailed comparison to both divertor effects like recycling or increased upstream SOL cross field transport.

Local Island Divertor Experiment

To achieve an improvement of plasma confinement by an effective edge plasma control, the local island divertor (LID) was originally proposed in the National Institute for Fusion Science in the early 1980s. The LID is a kind of island divertor that utilizes the island separatrix as the channeling magnetic structure of the divertor, and it has the particular characteristic of localizing the particle recycling in very small areas. Thus, it is possible to construct a compact closed divertor configuration with efficient pumping capability, which results in the low-recycling condition in the edge region. In this paper the LID project is reviewed, from the physics design phase with numerical validation or estimation of the LID principle to a recent experimental result of the superdense core mode, which is a promising discharge for next-generation devices.

The effects of an open and closed divertor on particle exhaust during edge-localized mode suppression by resonant magnetic perturbations in DIII-D

This paper compares the effects of divertor geometry on particle exhaust characteristics during the suppression of ELM using resonant magnetic perturbations (RMPs) on DIII-D. The subject is timely, particularly for ITER, because the combination of techniques to control or mitigate ELMs and control particle exhaust can provide confidence in the ability of an external pumping system to fully remove the particle exhaust. The differences between an open and closed divertor magnetic topology show a strong coupling of the perturbed strikepoint to the pumping manifold in closed divertor configurations, which can increase the particle exhaust by a factor of four. There is also an observed dependence on q95 in this configuration, which is a common feature of RMP ELM suppression. Neutral density in both the active and non-active divertors is seen to increase during the RMP in the ISS configuration, and edge plasma conditions (i.e. ne,sep and midplane profile of Dα) are seen to increase in the closed divertor configuration. Finally, the pumping exhaust is also shown to have a strong dependence on local measurements of the recycling flux. These observations, when taken as a whole, point to a substantial change in the plasma edge conditions, i.e. near the LCFS, throughout the poloidal cross-section of the vacuum vessel. This is coincident with the application of the RMP affecting the pumping capability of the system.

Steady-State Operation Scenario and the First Experimental Result on QUEST

QUEST focuses on the steady state operation of the spherical tokamak by controlled PWI and electron Bernstein wave current drive. One of the main purposes of QUEST is an achievement of long duration discharge with MW-class injected power. As the result, QUEST should be operated in the challenging region on heat and particle handling. To do the particle handling, high temperature all metal wall up to 623 K and closed divertors are planned, which is to realize the steady-state operation under recycling ratio, R = 1. This is a dispensable check to DEMO, because wall pumping should be avoided as possible in the view of tritium retention. The QUEST project will be developed in increment step such as, I. low β steady state operation in limiter configuration, II. low β steady state operation in divertor configuration, III. relatively high β steady state operation in closed divertor configuration. Phase I in the project corresponds to these two years, and final goal of phase I is to make full current drive plasma up to 20 kA. Closed divertor will be designed and tested in the Phase II. QUEST is running from Oct., 2008 and the first results are introduced.

Design of a Closed Helical Divertor in LHD and the Prospect for Helical Fusion Reactors

A new closed helical divertor configuration for efficient particle control and reduction of the heat load on the divertor plates is proposed. The closed divertor configuration practically utilizes an ergodic layer and magnetic field line configuration on divertor legs in helical systems. For optimization of the design of the closed divertor, the distribution of the strike points is calculated in various magnetic configurations in the Large Helical Device (LHD). It suggests that the installation of the closed divertor components in the inboard side of the torus under an inward shift configuration (Rax=3.60m) is the best choice for achieving the above two purposes. This divertor configuration does not interfere with plasma heating and diagnostic systems installed in outer ports. The prospect of the closed divertor configuration to a helical fusion reactor is investigated using a three-dimensional neutral particle transport simulation code with a one-dimensional plasma fluid calculation on the divertor legs. The investigation shows efficient particle pumping from the in board side and reduction of the heat load due to the combined effect of the optimized closed divertor geometry, ergodized divertor legs, and low electron temperature in the ergodic layer. It indicates a promising closed divertor configuration for helical fusion reactors.

Study of the effect of a closed divertor configuration on neutral particle control in the LHD plasma periphery

An optimized closed divertor configuration for effective particle control in LHD is proposed from the viewpoints of the distribution of the strike points and neutral particle transport. Calculations of the distribution of the strike points indicate that 50% of the strike points locate in the inboard side of the torus in a standard magnetic configuration ( R ax = 3.60 m). The ratio increases to 80% by installing target plates near lower/upper ports. A three-dimensional neutral particle transport simulation shows that installation of closed divertor components with the target plates raise the neutral pressure in the inboard side by more than one order of magnitude compared to that in the present open divertor case. The analysis of the neutral particle transport predicts that enhancement of the neutral pressure becomes moderate in outward shift configurations ( R ax > 3.75 m).

Septum assessment of the JET gas box divertor

The influence of the physical isolation of inner and outer divertor volumes by a septum plate of the Mk-II gas box divertor, thus increasing divertor closure and neutral compression, on the plasma and divertor performance has been studied at the Joint European Torus (JET). The septum plate was installed in 1999, together with the original Mk-II gas box divertor, and was then replaced by a simple protection plate in 2001. This removal reduced the closure of the divertor by opening a line of sight path for neutrals to travel between the inner to the outer divertor volumes. Comparison of identical discharges with and without the septum thus provides direct evidence of the effect of divertor closure on plasma behaviour. With this aim, following septum removal, several dedicated L-mode and H-mode discharges have been performed, in each case repeating earlier discharges when the septum was still in place. In each case, the fuelling location was varied between the inner/outer divertor and the main chamber, and differences in detachment in the inner and outer divertors were studied. Under L-mode conditions, differences in detachment dynamics were indeed observed between closed (with septum) and open (without septum) divertor configurations, although the differences were only significant in the medium density range. In contrast, the ultimate density limit was not affected, being determined in each case by the formation of a wall multifacedted asymmetric radiation from the edge (MARFE), rather than an X-point MARFE. Under H-mode conditions, the differences were more subtle. Although the ion fluxes to the targets were unaffected, the target electron temperatures were found to be lower in the closed divertor configuration. In this case, the fuelling efficiency was the largest when the gas injected from the inner divertor, with implications on global energy confinement and ELM frequency. Otherwise, no difference in the confinement of the discharges with and without septum was observed.

Optimization of Closed Divertor Configuration for Effective Particle Control in LHD Plasmas

AbstractNeutral particles in the LHD plasma periphery are investigated by three‐dimensional neutral particle transport simulation. The simulation predicts significant change of the density profile of neutral particles depending on magnetic configurations (radial position of the magnetic axis Rax ). The polarization resolved Hα spectra and the Hα intensity profile measured in the three different magnetic configurations are consistent with calculations by the simulation. It shows that the density of neutral hydrogen molecules in the inboard side of torus is relatively high in magnetic configurations (3.50m < Rax < 3.75m), which also indicates that enhancement of the neutral density by more than one order of magnitude is necessary for effective particle control in LHD plasmas. The simulation strongly suggests that installation of closed divertor components in the inboard side is reasonable. The simulation coupled with a one‐dimensional plasma fluid analysis on divertor legs proposes an optimized closed divertor configuration which can enhance the neural density in the inboard side of the torus. (© 2008 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Study of a Closed Divertor Configuration for the Three-Dimensionally Complicated Magnetic Structure in the LHD Plasma Periphery

The distribution of strike points in a closed divertor configuration with additional baffle plates installed in the toroidal ends of closed divertor components near lower/upper ports is calculated by tracing magnetic field lines from the last closed flux surface in various magnetic configurations (radial position of the magnetic axis, Rax = 3.50 ∼ 3.90 m). The calculation shows that the ratio of the number of strike points on additional plates is maximum for Rax = 3.60 m, showing that the plates are effective in this magnetic configuration. Neutral particle transport is investigated using a fully three-dimensional code EIRENE with one-dimensional plasma fluid analysis of divertor legs, where the plasma parameter profiles on the legs are obtained by an iterative calculation including interaction processes between the plasma and neutral particles. The plates raise the pressure of molecular hydrogen locally near the baffle plates to more than 0.2 Pa, which is enough for efficient particle control using vacuum pumping systems installed near the plates. The simulation proposes one possible candidate of optimized closed divertor configuration for the three-dimensionally complicated magnetic structure in the LHD plasma periphery.

Engineering design and thermal hydraulics of plasma facing components of SST-1

SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (≈1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2. In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 °C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a fortran code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ansys software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper.

Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up–down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m 2. In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350°C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper.

Recycling of gaseous impurities in ASDEX

The closed divertor configuration of ASDEX isolates the zone of dominant plasma-wall interaction in a rather well defined manner from the plasma in the main chamber. Studies have shown that for recycling gases the effective conductance between divertor and plasma chamber is closely represented by the values for molecular flow also during a discharge. Hence, from partial pressure measurements of contaminant gases in the divertor impurity recycling fluxes can be obtained. In deuterium discharges the problem of mass peak interferences, especially for methane and watervapour, has to be resolved. Data are shown for various ASDEX scenarios: stainless steel walls, carbon wall elements, Ti-gettering and boronization. The results expose the production of CO as main culprit, as long as no gettering or boronization is employed. Then, however, with carbon still present in the machine, the hydrocarbons limit the attainment of optimum performance parameters. What are the conclusions?

Recycling and neutral transport in the DIII-D tokamak

Several diagnostics have been used to characterize the edge plasma in the DIII-D divertor region and thereby to understand the particle recycling and neutral particle transport in an open divertor. An array of Langmuir probes mounted on the carbon divertor plates determines that generally the electron temperature is low ( T e ⋍ 10–20 eV ), and the electron density is high ( n e ⋍ 5 × 10 19 m −3 ). The edge plasma temperature and density are also measured by a moveable Langmuir probe mounted at the midplane of the plasma; the data from these shot-by-shot edge radial scans are then connected with the data for the core plasma obtained by Thomson scattering. The heat flux to the divertor plates is measured by an absolutely calibrated infrared camera; a plasma model is used to compare the heat flux with the measured T e and n e . The molecular neutral pressure at the edge of the plasma at several poloidal and toroidal locations is obtained from absolutely-calibrated pressure gauges; a gas puff enables in-situ gauge calibrations before each plasma shot. Typically, the divertor pressure is 10–50 times larger than that at the midplane. Time-resolved H α brightness measurements are obtained from an absolutely calibrated television camera viewing the divertor region from above; both strike points are viewed simultaneously. The emission from the inner and outer strike points are usually equal after the H-mode transition, but are often unequal after the first ELM in H-mode and during some phases of L-mode discharges. A tangential camera measures the falloff in emission from the X-point to the plasma midplane. These measurements have been modeled with the DEGAS neutral transport code specifically modified for DIII-D. The wall geometry, wall composition, measured magnetic geometry, and measured plasma parameters are inputs to the model. The code calculates the gauge pressure as a function of position and the H α television picture directly. During H-mode, we find a factor of two agreement in both the measured pressures and the absolute H α brightnesses with one exception: the measured divertor pressure is higher than the code prediction. However, we find that a local gas source located below the X-point equal to only 10% of the divertor recycling source will bring the divertor pressure into agreement with the data. A difference in the electron temperature and density at the two strike points has been used to model the asymmetric H α emissions. Comparisons have been made between the model and the data during H-mode periods of the discharge. The model has also been used to predict the influence of baffles to form a more closed divertor configuration.