Divertor Plasma Temperature Research Articles

Tungsten divertor sources in WEST related to impurity inventory and local plasma conditions

A dedicated series of L-mode deuterium discharges were performed in WEST in which a broad parameter space in terms of divertor plasma temperature (inner divertor at <10 eV, outer divertor up to 50 eV) and impurity flux density was achieved by variation of X point elevation, upstream plasma density and fuelling position and rate. Density steps provided stable reference plasma conditions whilst density scans up to the point of detachment revealed the sputtering threshold behaviour of the dominant sputtering species. Tungsten gross erosion was quantified via the common WI 400.9 nm line and correlated with the residual gas content.

Simulations on W impurity transport in the edge of EAST H-mode plasmas

This paper provides an investigation of the production/transport properties of W impurity in the edge of experimental advanced superconducting tokamak (EAST) H-mode discharges with upper-single-null configurations by using DIVIMP Monte-Carlo code. The background plasmas are provided by SOLPS5.0 calculations. Firstly, to address the detailed dependence of W impurity behaviors on plasma conditions in the scraped-off layer (SOL)/divertor region of EAST, two pre-edge localized mode (ELM) cases with divertor plasmas in high-recycling and partially-detached regimes, respectively, are considered and compared. It is found that, due to the competitions between the thermal and friction forces (FFs), in the high-recycling case, large quantities of W impurities are transported to the upstream; while in the partially-detached case, most of the W impurities are located near the inner and outer divertors. Furthermore, the W core contamination in type-I ELMy H-mode plasmas of EAST has been estimated by the DIVIMP-SOLPS5.0 simulations, without the consideration of the release of W impurity from the core due to ELMs. During the ELM, simulations exhibit the low-recycling regime of divertor plasma, and high (several hundred eV) target plasma temperatures together with low SOL collisionality ( ∼ 1–3), lead to an order of magnitude increase in the W core contamination rate. However, in the ELM-recovering phase, with the decrease (increase) of divertor plasma temperature (density), significant increase of divertor retention is obtained due to the remarkable increase of FFs on W ions. Besides, intra-ELM phase approximately contributes to more than 50% of the total W core contamination per an ELM cycle. This work represents a step toward a deeper understanding of the W impurity production/transport in EAST H-mode discharges.

ITER upper port wide angle viewing system optical design and performance analysis

The Upper port Wide Angle Viewing System (UWAVS) is diagnostic system that provides visible light and InfraRed images of the divertor area in the ITER tokamak. The main functions of the UWAVS are to provide divertor wall and plasma luminance and temperature distributions at high spatial resolution (3–5 mm) and frame rates (500–1000 Hz) for machine protection as well as to enable species influx monitoring (Be, C, O, Ne, Kr, Ar) to the divertor. In addition the UWAVS supports visual inspection as well as various other measurements. For this purpose an optical system is designed to image the divertor area onto four different cameras in the port cell area. Altogether five of these systems are foreseen to provide sufficient coverage of the divertor area (150m2). Each of these systems is equipped with active alignment to compensate for vessel motion, internal transmission monitoring sources to assess mirror reflectivity degradation, as well as calibration sources in the port cell to monitor the camera pixel response, all to guarantee the overall systems performance during operation and over lifetime. All these aspects are combined into an optical design targeting diffraction limited imaging, taking in account the complexities of the harsh ITER environment. This design and its performance are supported by manufacturability analysis, tolerancing analysis and alignment strategy and Structural Thermal Optical Performance (STOP) analysis. The latter is an TNO in-house developed method that provides a direct and reproducible link between the structural thermal analysis for a combined load case comprising nuclear heating, cooling water pressure and temperature, gravity and mounting stresses (in Ansys) and the associated changes of the optical performance, including imaging quality and alignment (in Zemax OpticStudio).

Constraining the divertor heat width in ITER

A model is developed which constrains heat width, λr based on global power balance, momentum conservation, pedestal stability and sheath heat transmission. The model relies on measurements of the ratio of separatrix to pedestal pressure; a ratio ∼5% is found to be expected for ITER. Applying this model indicates a constraint that the allowed λr∼10–30mm for ITER if the divertor is in the high-recycling regime as expected (T<20eV) while a λr∼1–3mm requires a separatrix pressure approximately equal to the top pedestal pressure in violation of physical reasoning and the concept of a pedestal. A weaker constraint is applied in the model that upstream separatrix temperature simultaneously satisfies power balance. The constrained model cannot satisfy power balance with λr<3mm, and in order to obtain λr∼5mm requires divertor plasma temperature >100eV, a condition which would have very negative consequences for the divertor, but has never been observed experimentally.

Tungsten divertor erosion in all metal devices: Lessons from the ITER like wall of JET

Tungsten erosion in the outer divertor of the JET ITER like wall was quantified by spectroscopy. Effective sputtering yields of typically 10−4 were measured in L-mode at ∼30eV attached divertor conditions and beryllium was identified as the main cause of sputtering. The signature of prompt redeposition was observed in the analysis of WI 400.9nm and WII 364nm line ratios and indicative of >50% redeposition fractions. Inter- and intra-ELM sputtering were compared for an example of 10Hz ELMs with 13MW NBI heating, in which intra-ELM sputtering was found to dominate by a factor of 5. Nitrogen seeding initially increased the tungsten sputtering threefold due to higher extrinsic impurity levels and effectively reduced the tungsten sputtering when the divertor plasma temperature was decreased from the initial 25eV down to 15eV.

First study of EAST divertor by impurities puffing

A series of experiments has recently been carried out in EAST under different plasma conditions to investigate the basic divertor performance, divertor and SOL screening efficiency and radiative divertor effect. Detached divertor plasmas have been achieved by density ramp-up. It is found that CIII emission from the low-field side (LFS) exhibits a strong dependence on poloidal locations and plasma operation regimes from methane (CH4) puffing experiments. In addition, the radiative divertor experiments by injection of mixed Ar (5.7% Ar in D2) into the outer divertor chamber reduce the peak heat flux by 50% at the outer target plate, which also reduce the divertor plasma temperature. The in–out heat flux distribution asymmetry is improved.

The super X divertor (SXD) and a compact fusion neutron source (CFNS)

A new magnetic geometry, the super X divertor (SXD), is invented to solve severe heat exhaust problems in high power density fusion plasmas. SXD divertor plates are moved to the largest major radii inside the TF coils, increasing the wetted area by 2–3 and the line length by 2–5. Two-dimensional fluid simulations with SOLPS (Schneider et al 2006 SOLPS 2-D edge calculation code Contrib. Plasma Phys. 46) show a several-fold decrease in divertor heat flux and plasma temperature at the plate. A small high power density tokamak using SXD is proposed, for either (1) useful fusion applications using conservative physics, such as a component test facility (CTF) or fission–fusion hybrid, or (2) to develop more advanced physics modes for a pure fusion reactor in an integrated fusion environment.

Modelling of 13CH4 injection and local carbon deposition at the outer divertor of ASDEX Upgrade

Numerical modelling of 13CH4 injection into the outer divertor plasma of the full tungsten, vertical target of ASDEX Upgrade is presented. The SOLPS5.0 code package is used to calculate a realistic scrape-off layer plasma background corresponding to L-mode discharges in the attached divertor plasma regime. The ERO code is then used for detailed modelling of the hydrocarbon break-up, re-deposition and re-erosion processes. The deposition patterns observed at two different poloidal locations are shown to strongly reflect the cross-field gradients in divertor plasma density and temperature, as well as the local plasma collisionality. Experimental results with forward and reversed BT, accompanied by numerical modelling, also point towards a significant poloidal hydrocarbon E×B drift in the divertor region.

Electron density measurements in the National Spherical Torus Experiment detached divertor region using Stark broadening of deuterium infrared Paschen emission lines

Spatially resolved measurements of deuterium Balmer and Paschen line emission have been performed in the divertor region of the National Spherical Torus Experiment using a commercial 0.5m Czerny-Turner spectrometer. While the Balmer emission lines, as well as the Balmer and Paschen continua in the ultraviolet and visible regions have been extensively used for tokamak divertor plasma temperature and density measurements, the diagnostic potential of infrared Paschen lines has been largely overlooked. We analyze Stark broadening of the lines corresponding to 2−n and 3−m transitions with principal quantum numbers n=7–12 and m=10–12 using recent model microfield method calculations [C. Stehle and R. Hutcheon, Astron. Astrophys., Suppl. Ser. 140, 93 (1999)]. Densities in the range (5–50)×1019m−3 are obtained in the recombining inner divertor plasma in 2–6MW neutral beam heated H-mode discharges. The measured Paschen line profiles show good sensitivity to Stark effects and low sensitivity to instrumental and Doppler broadenings. The lines are situated in the near-infrared wavelength domain, where optical signal extraction schemes for harsh nuclear environments are practically realizable and where a recombining divertor plasma is optically thin. These properties make them an attractive recombining divertor density diagnostic for a burning plasma experiment.

Asymmetry of dense divertor plasmas influenced by thermoelectric potential and charge-exchange momentum loss

Asymmetric equilibria of divertor plasmas influenced by the thermoelectric potential and the momentum loss due to the charge-exchange collisions have been studied by using a five-point model. The momentum loss becomes large when the divertor plasma temperature decreases below about 10 eV. The increase of the pre-sheath potential due to the momentum loss compensates the reduction of the sheath potential at the low-temperature side of an asymmetric equilibrium. Through the change of the pre-sheath potential, the momentum loss has a stabilizing effect on the thermoelectric instability and mitigates the asymmetry at the high-density divertor regime. The momentum loss at the low-temperature and high-density side divertor is effective to lower the temperature at the high-temperature side divertor.

Ergodic divertor experiments on the route to steady state operation of Tore Supra

Ergodic divertor operation on Tore Supra is characterized by good performance in terms of divertor physics. Control of particle recirculation and impurity screening are related to the symmetry, both poloidally and toroidally, of the shell of open field lines and to its radial extent, Δr ≈ 0.16 m. Feedback control of the divertor plasma temperature has led to controlled radiative divertor experiments. In particular, good performance is obtained when the plasma is controlled to be at a temperature comparable to the energy involved in the atomic processes (15-20 eV). For standard discharges with 5 MW total power and ICRH heating, the low parallel energy flux ≈ 10 MW m-2 is reduced to ≈ 3 MW m-2 with nitrogen injection. This is achieved at a modest cost in core dilution, ΔZeff ≈ 0.3. Despite the large volume of open field lines ( ≈ 36%), the ergodic divertor does not reduce the possible current in the discharge since stable discharges are achieved with qsep ≈ 2. It is shown that the reorganization of the current profile in conjunction with a transport barrier in the electron temperature on the separatrix stabilizes the (2,1) tearing mode. Confinement follows the standard L mode confinement. In a few cases at high density and with no gas injection (wall fuelled discharges), `RI-like' modes are reported with a modest increase in confinement ( ≈ 40%). Despite the lack of core fuelling on Tore Supra, high densities during ICRH pulses can be achieved with Greenwald fractions fG ≈ 1. Compatibility with both ICRH and LH is demonstrated. In particular, long pulse operation with a flat-top in excess of 20 s is achieved with LHCD.

Numerical analysis of thermoelectric instability in tokamak divertor

The thermoelectric instability in tokamak divertor has been studied numerically by using a five-point model. When the divertor plasma temperature T div decreases below a critical temperature, this instability arises in a symmetric equilibrium of scrape-off layer (SOL) and divertor plasmas, and stable asymmetric equilibria are found. The SOL current causes spontaneously the divertor asymmetry at the onset of the thermoelectric instability. In an asymmetric equilibrium, the SOL current flows from the higher- T div side to the lower- T div side. The heat flux flowing into the divertor plate of the higher- T div side becomes larger than that of the lower- T div side. The heat flux asymmetry is larger for cases of high recycling and low heating power.

Onset condition of thermoelectric instability in divertor tokamaks

The onset condition of a thermoelectric instability in divertor tokamaks has been studied analytically and numerically by using a five point model based on SOL and divertor plasmas. The SOL current due to the thermoelectric effect causes the thermoelectric instability and destabilizes the symmetric equilibrium of the SOL and divertor plasmas below a critical divertor plasma temperature. The critical temperature of the thermoelectric instability is equal to or higher than that of the thermal instability due to the divertor radiation, and depends on the midplane SOL plasma temperature and the divertor radiation loss. Below the critical temperature, stable asymmetric equilibria are obtained. As the degree of the asymmetry in the divertor plasma temperature increases, the SOL current has no effect on the instability, or even stabilizes the equilibrium. The SOL current generates spontaneously the divertor asymmetry through the thermoelectric instability.

A numerical study of density limit mechanisms in tokamaks

A core particle and power balance model with radiative cooling, a ballooning-mode-limit pedestal model, a “two-point” plasma divertor model with radiative and atomic recycling cooling, and a two-dimensional (2-D) neutral transport model have been combined to model the coupled plasma core-edge physics in a tokamak. This model has been applied to examine the physical mechanisms which limit the attainable core density. The first limitation is fueling rate and core penetrability. For fueling that is able to penetrate the core, the core density increases with a fueling rate up to a limit set by one of two other mechanisms. When a sufficient impurity concentration is present, the mechanism that limits the maximum density attainable in the core seems to be an impurity-driven thermal instability that produces a thermal collapse of the core plasma. At lower impurity concentrations, the buildup of plasma density in the divertor and the associated increase in atomic recycling and ionization cooling causes a collapse of the divertor plasma temperature, leading to a thermal collapse of the core plasma. An increase in core heating input power or in plasma confinement extends the point of thermal collapse to higher density.

Power plant design study of a high aspect ratio Tokamak using a SiC composite structure

The DREAM (DRastically EAsy Maintenance) Tokamak is a fusion power plant which is designed from the viewpoint of maintenance feasibility. For this purpose, the DREAM reactor uses a plasma with a very high aspect ratio ( A) and adopts SiC as a structural material. The choice of SiC affects the design of the core plasma, i.e. large inboard shield thickness, low synchrotron radiation reflectivity, and small plasma elongation for positional stability. The objectives of this study are to explore the feasibility of a high- A device, such as a power plant, and to clarify the technological impact of SiC material on the plasma design. Plasma size is optimized by the physics guidelines similar to ITER. The plasma major and minor radii of DREAM are 16 m and 2 m, respectively, and the average neutron wall load is 2.5 MW m −2, the maximum toroidal field is 20 T, and the fusion power is 5.5 GW. Steady-state operation is obtained with 50 MW of external current-drive power and 90% bootstrap current. The divertor heat load is estimated to be about 10 MW m −2. A radiative divertor concept is adopted to achieve a low divertor plasma temperature. The DREAM Tokamak concept is found to be a possible candidate for a future power plant with more than 5 GW of fusion power and an acceptable divertor condition.

The two-dimensional structure of radiative divertor plasmas in the DIII-D tokamak

Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne⩽1021 m−3, 0.5 eV⩽Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion–neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3–5, consistent with the measurements. Absolutely calibrated vacuum ultraviolet spectroscopy and 2-D images of impurity emission show that carbon radiation near the X-point, and deuterium radiation near the target plates contribute to the reduction in Te. Uniformity of radiated power (Prad) (within a factor of 2) along the outer divertor leg, with peak heat flux on the divertor target reduced fourfold, was obtained. A comparison with 2-D fluid simulations shows good agreement when physical sputtering and an ad hoc chemical sputtering source (0.5%) from the private flux region surface are used.

Access conditions for H modes with detached divertor plasmas

Access conditions for H modes with detached divertor plasmas are derived by considering power threshold, divertor plasma temperature and density limit constraints. It is found that there is a minimum scrape-off layer power and main plasma density for accessing the detached divertor H mode, and that the required divertor radiated power fraction must increase as the machine size and field increase. The model describes, at least qualitatively, experimental findings on ASDEX Upgrade, DIII-D, JET and JT-60

Review of divertor diagnostics (invited) (abstract)a)

We will review tokamak divertor diagnostics currently in use to study issues relevant to next generation divertor design. The divertor represents perhaps the greatest technological challenge in the design of reactor tokamaks such as ITER, with designs currently requiring a reduction in heat flux to plasma facing surfaces obtained through radiative and charge exchange processes (i.e., the radiative divertor). Experiments to study these new divertor regimes and concepts require 2D profile measurements of the important plasma parameters, scrape-off-layer (SOL) and divertor plasma density and temperature, radiated power, impurity concentrations, particle sources, neutral pressure, and heat flux to surfaces. Diagnostics measurements in use on ASDEX-U, Alcator C-MOD, DIII-D, JET, and JT-60U include Langmuir probes, IRTV cameras, bolometer arrays, fast neutral pressure gauges, photodiode arrays, CCD cameras, and visible and VUV spectroscopy. Each diagnostic obtains unique information, but each also presents its own challenges of access and interpretation. For the future, ITER divertor diagnostics will require innovation to adapt to the long pulse, high power, and neutron environment.

A first flight corrected neutral diffusion model for edge plasma simulation

The standard neutral diffusion model is improved by treating the first flight neutrals, originating at the boundary, separately and exactly, and using the diffusion approximation only for the neutrals suffering one or more charge exchange interactions. Different source neutral energies are thereby accomodated unlike the diffusion model which cannot distinguish between bulk plasma neutrals and recycling neutrals. Both models have been implemented in the 2D NEWEDGE code and results for 5 and 50 eV neutrals are contrasted with diffusion model predictions. Significant differences are seen in the divertor plasma density, temperature and heat flux profiles, with the 50 eV neutral energy case showing most dramatic differences. An improvement in convergence rate of the edge plasma code is also seen for the first flight corrected diffusion model.

Divertor retention for recycling impurities

A quick and simple experimental method of comparing the global retention capabilities of different divertor configurations for gaseous impurities is presented. Short gas puffs into the main chamber and analysis of the time dependence of central impurity lines are used to determine the retention capability in terms of time constants, which characterize the impurity leakage rate from the divertor chamber to the main plasma and to the pumping system. A simple two-reservoir model is used to define these basic parameters of divertor retention. For a more detailed evaluation of the divertor retention parameters from spectroscopic data, an impurity transport code is used to simulate the impurity radiation measured in various experiments. The dependence of the retention capability on the divertor plasma density and temperature as well as on the distance along the field lines back to the main plasma is discussed. The dependence on the impurity species is also investigated. On the basis of these results, the consequences of the divertor retention for helium exhaust in future fusion devices are analysed by numerical simulation