Scrape Off Research Articles

Role of radiation re-absorption in the thermal helium beam diagnostic.

The Thermal Helium Beam (THB) is a diagnostic for simultaneously measuring the electron temperature and density profiles of the plasma edge and scrape off layer (SOL). It exploits the line ratio technique of selected He line intensities, emitted by He gas puffed inside the plasma, to locally estimate the plasma properties through a dedicated collisional radiative model (CRM). Standard THB diagnostics used in nuclear fusion devices measure three HeI emission lines: 667.8, 706.5, and 728.1nm. For the RFP experiment RFX-mod2, a new THB is designed and tested for the first time at the TCV tokamak. It acquires an additional emission line at 501.6nm, which is exploited to estimate the radiation re-absorption, which is not negligible in regions of large neutral He densities (leading to high re-absorption) and simultaneously low electron density and temperature (lack of other excitation channels). It affects the measurements most strongly at the far SOL, while the significance of re-absorption decreases as it approaches the separatrix. In this paper, plasma density and temperature profiles of the plasma edge at the outboard midplane of TCV, measured with this newly designed THB, are presented. For the first time, the effect of radiation re-absorption on the estimation of electron temperature and density profiles is experimentally measured in a tokamak using the 501nm line emission intensity. Different CRMs are compared with and without radiation re-absorption, showing good agreement when re-absorption is included and demonstrating how it plays an important role in the far SOL, as expected.

Design of a low frequency, density profile reflectometer system for the MAST-U spherical tokamak.

Validated and accurate edge profiles (temperature, density, etc.) are vitally important to the Mega Ampere Spherical Tokamak Upgrade (MAST-U) divertor and confinement effort. Density profile reflectometry has the potential to significantly add to the measurement capabilities currently available on MAST-U (e.g., Thomson scattering and Langmuir probes). This work presents the diagnostic requirements, problems, and solutions facing profile reflectometry in spherical tokamaks and MAST-U in particular. Requirements include density measurements near zero electron density in the scrape off layer region, coverage for a broad range of MAST-U plasma parameters, high time (≤10microseconds) and spatial resolutions (≤1cm), reliability, and identification of the plasma start frequency.

Calorimetry measurement for energy balance and energy distribution in WEST for L-mode plasmas

This paper presents the energy balance of 602 pulses from four different experimental campaigns for the WEST tokamak. Different magnetic configurations have been studied, with lower single null (LSN) and upper single null (USN) configuration with deuterium or helium plasmas. The energy balance is closed with an imbalance of about 5% of the total injected energy for most of the campaigns and for different magnetic configurations. The distribution over the whole machine is shown, with the outer first wall receiving most of the energy due to its large surface area with about 30% of the total heat load, and the divertor with 25% due to the heat loads deposited by the convected power in the scrape-off layer (SOL). Finally, the tomography inversion of the bolometry measurement allows us to disentangle the contribution of the radiated and convected power in the energy absorbed by each type of plasma-facing component. We show that in the USN configuration about 63% of the available energy in the SOL is deposited in the upper divertor (UDIV) through convected heat loads, while in LSN this value is spread over the lower divertor with 45% and the baffle and UDIV with about 10% for both.

The observation of avalanche-like transport during confinement power degradation in Heliotron J

In this study, we observe avalanche-like electron thermal transport in Heliotron J under plasma conditions that exhibit power degradation to the central electron cyclotron heating (ECH). The newly installed GHz sampling electron cyclotron emission diagnostic allows for the observation of the long-distance radial propagation of electron temperature fluctuations. We find that the electron temperature fluctuations are associated with avalanches, because the Te fluctuations (i) propagate from the core to the edge with a speed comparable to the diamagnetic drift velocity, (ii) exhibit a 1/f power-law scaling in the frequency spectrum and a Hurst exponent close to 1, and (iii) dominate in the ECH deposition location and spread to the edge as the heating power increases. Furthermore, the electron heat avalanches can spread to the scrape off layer region when they are enhanced, i.e., the Te fluctuations correlate with the Dα emission, which also has a frequency spectrum that scales with a 1/f power law.

The effect of gas injection location on a lithium vapor box divertor in NSTX-U

The lithium vapor box divertor is a proposed divertor design that will minimize contamination of the upstream plasma in a fusion device, while also ensuring protection of the target. In this design lithium is evaporated near the target by high temperature lithium surfaces, dissipating the plasma heat flux. The lithium vapor box has been predicted via the fluid-kinetic code scrape off layer plasma simulator (SOLPS-ITER) to achieve low (/0.05) upstream concentrations of lithium and low target heat fluxes. Here we compare two choices of deuterium gas puff location using SOLPS-ITER, the private flux region (PFR) and the common flux region (CFR), and find significant differences in the contamination level required to reach an acceptable target heat flux (defined here as q10 MW m−2). Deuterium gas puffing from the PFR is seen to better reduce upstream lithium contamination. The difference in puffing location is seen to cause changes in the upstream flow of lithium ions. The PFR puff, having better access to the separatrix, can better reduce the upstream-directed flow of lithium near the separatrix which is the primary source of contamination due to a large thermal force in this region. Puffing from the CFR, partially due to inefficacy at reducing separatrix lithium flow, has higher lithium concentration within the plasma. Solutions that reduce the heat flux to below 10 MW m−2 have a range of lithium concentrations between /0.01–0.12 depending on puff intensity, location, evaporator temperature and recycling at the various plasma facing components. The efficacy of the puffs is tested for sensitivity to deuterium recycling coefficient at the target, evaporators, and main chamber walls. A CFR located puff is found to be more dependent on the recycling coefficients used than a PFR located puff. regardless of the set of recycling coefficients chosen, PFR puffing achieves lower lithium contamination than CFR puffing for a given heat flux.

Modeling snowflake plasmas in RFX-mod2: A test bed for SOL and edge physics characterization

The definition of low-current snowflake plasma equilibria in RFX-mod2 experiment is carried out in terms of both plasma conditions and engineering requirements. The snowflake magnetic configuration is obtained by solving a least-squares curve fitting problem with properly defined linear contraints. Positive and negative triangularity snowflake equilibria are defined and compared with a standard single-null plasma. The magnetic field topology is characterized in terms of local magnetic shear and connection length revealing all the potential advantages of a wider low poloidal field region. The RFX-mod2 improved diagnostic system and its magnet flexibility would allow a detailed characterization of the scrape off layer and plasma edge physics in snowflake configurations.

Simulation of Liquid Lithium Divertor Geometry Using SOLPS-ITER

Plasma-facing component (PFC) geometries are evaluated for a Fusion Nuclear Science Facility (FNSF) design to find solutions that are compatible with flowing liquid lithium components while satisfying requirements from the perspective of plasma and neutral particle transport. Flowing liquid metal (LM) divertor systems offer important advantages due to continual regeneration of the surface material, but may require modifications from standard optimized tokamak divertor designs. Scrape Off Layer Plasma Simulator for the International Thermonuclear Experimental Reactor (SOLPS-ITER) simulations are used to compare standard vertical target, open, and balanced baffled geometries. It is found that the open geometry offers little control of neutral particles and has narrow operating windows where the divertor fluxes can be lowered to acceptable levels, while maintaining the required upstream density. The addition of neutral baffles allows greater control over the divertor conditions with less impact on the upstream density. The divertor neutral pressure in the baffled geometry is slightly higher than the open divertor, but not as large with a vertical target. A first-pass coupling of the SOLPS solution with sheath and implantation/erosion models to compute the lithium emission shows that the lithium is largely confined to the near-surface region due to the small ionization mean free path and large main ion flow toward the surface. The lithium source was varied by three orders of magnitude. At the highest rate, when the sourced lithium approaches the level of the main ion recycling flux, a high core lithium concentration (~10%) is observed with only moderate power dissipation to affect the divertor plasma.

Development of two dimensional full wave spectral code for the ICRF heating and current drive research including scrape-off layer in tokamaks

It is important for an ICRF full wave code to simulate the SOL (Scrape Off Layer) plasma as well as the core inside of the LCFS (Last Closed Flux Surface) for the precise prediction of the coupling between the antenna and the core plasma in tokamaks. To this end, a two dimensional full wave code based on a Fourier spectral algorithm has been developed. The spectral algorithm and procedures are described and the simulation results for the minority heating in KSTAR are reported including electric field, power absorption and power flux.

Characteristics of Density and Temperature Fluctuation in Fusion Edge Plasma and Implication on Scrape off Layer Width

Characteristics of Density and Temperature Fluctuation in Fusion Edge Plasma and Implication on Scrape off Layer Width

Dashnaktsutyun and its anti-human propaganda

Dashnaktsutyun and its anti-human propaganda

Statistical properties of ion and electron temperature fluctuations in the edge of the COMPASS tokamak

This paper presents measurements of the plasma turbulence with an emphasis on the ion temperature fluctuations in comparison to the electron temperature and the ion saturation current fluctuations in the edge of the COMPASS tokamak. The ion temperature values are obtained by using a fast swept ball-pen probe, with temporal resolution up to 10 µs, accompanied by independent 1 µs measurements of the electron temperature and ion saturation current. The experimental results obtained for a wide range of line-averaged densities show that statistical properties of the fluctuations are in good agreement with a form of beta or gamma distribution function. In addition, the dependence of the averaged ion and electron temperatures on the line-averaged density is studied in the near and main scrape off layer region.

Non-linear MHD modelling of edge localized modes suppression by resonant magnetic perturbations in ITER

Edge localized modes (ELMs) suppression by resonant magnetic perturbations (RMPs) was studied with the non-linear magneto-hydro-dynamic (MHD) code JOREK for the ITER H-mode scenarios at 15 MA, 12.5 MA, 10 MA/5.3 T. The main aim of this work was to demonstrate that ELMs can be suppressed by RMPs while the divertor 3D footprints of heat and particle fluxes remain within divertor material limits. The unstable peeling–ballooning modes responsible for ELMs without RMPs were modelled first for each scenario using numerically accessible parameters for ITER. Then the stabilization of ELMs by RMPs was modelled with the same parameters. RMP spectra, optimized by the linear MHD MARS-F code, with main toroidal harmonics N = 2, N = 3, N = 4 have been used as boundary conditions of the computational domain of JOREK, including realistic RMP coils, main plasma, scrape off layer (SOL) divertor and realistic first wall. The model includes all relevant plasma flows: toroidal rotation, two fluid diamagnetic effects and neoclassical poloidal friction. With RMPs, the main toroidal harmonic and the non-linearly coupled harmonics remain dominant at the plasma edge, producing saturated modes and a continuous MHD turbulent transport thereby avoiding ELM crashes in all scenarios considered here. The threshold for ELM suppression was found at a maximum RMP coils current of 45 kAt–60 kAt compared to the coils maximum capability of 90 kAt. In the high beta poloidal steady-state 10 MA/5.3 T scenario, a rotating QH-mode without ELMs was observed even without RMPs. In this scenario with RMPs N = 3, N = 4 at 20 kAt maximum current in RMP coils, similar QH-mode behaviour was observed however with dominant edge harmonic corresponding to the main toroidal number of RMPs. The present MHD modelling was limited in time by few tens of ms after RMPs were switched on until the magnetic energy of the modes saturates. As a consequence the thermal energy was still evolving on this time scale, far from the ITER confinement time scale and hence only the form of 3D footprints on the divertor targets can be indicated within this set-up. Also note, that the divertor physics was missing in this model, so realistic values of fluxes are out of reach in this modelling. However the stationary 3D divertor and particle fluxes could be simply extrapolated from these results to the stationary situation considering that a large power fraction should be radiated in the core and SOL and only about 50 MW power is going to the divertor, which is an arbitrary, but reasonable number used here. The 3D footprints with RMPs show the characteristic splitting with the main RMP toroidal symmetry. The maximum radial extension of the footprints typically was ∼20 cm in inner divertor and ∼40 cm in outer divertor with stationary heat fluxes decreasing further out from the initial strike point from ∼5 MW m−2 to ∼1 MW m−2 assuming a total power in the divertor and walls is 50 MW. The heat fluxes remain within the divertor target and baffle areas, however with rather small margin in the outer divertor which could be an issue for the first wall especially in transient regimes when part of the plasma thermal energy is released due to switching on the RMP coils. This fact should be considered when RMPs are applied with a more favorable application before or soon after the L–H transition, although optimization is required to avoid increasing the L–H power threshold with RMPs.

Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared thermography, thermal measurement with thermocouples and fibre Bragg grating embedded few mm below the surface and flush mounted Langmuir probes. A large database including different magnetic equilibrium and input power is investigated to compare the heat load pattern (location, amplitude of the peak and heat flux decay length) on the inner and outer strike point regions: from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (<1 cm) is obtained between thermal inversions and flush-mounted LP measurements. The peak heat flux from the whole set of diagnostics is in good agreement and mainly in the ±20% range, while the heat flux decay length reported on the target shows significant discrepancy between diagnostics and location in the machine (±40% range). Despite such discrepancy, heat flux decay length at target is found to scale mainly with the magnetic flux expansion through the variation of the X-point height, as expected. The improved plasma performances achieved during C4 enabled to reach significant heat load in the divertor, up to 6 MW m−2 with 4 MW of additional heating power showing the capability to reach the ITER relevant heat load (10 MW m−2 steady state) with about 7 MW of additional power in L-mode discharge. The heat load distribution is clearly asymmetric with a 3/4 and 1/4 distribution on the outer and inner strike point region respectively for the parallel heat flux.

The effects of E r × B t drift on LH wave coupling with divertor configuration at EAST

The coupling characteristics of the 4.6 GHz lower hybrid (LH) grill and the factors affecting LH wave coupling, mainly drift, with divertor configuration in anti-clockwise/clockwise magnetic field direction are studied. Based on the Langmuir probe measurement, the relationship between the reflection coefficient (RC) of LH power and the edge density at the grill mouth () measured by Langmuir probes is investigated. The optimum LH wave coupling with RC less than 2% is achieved when is in the range of –, which is quite close to the theoretically predicted value. In addition, the experimental results show that drift in the divertor scrape off layer plays a dominant role in affecting and LH wave coupling. Studies show that the discrepancy in and LH wave coupling in different magnetic field directions is mainly attributed to the different directions of drift, whereas the corresponding discrepancy with different plasma currents is mainly ascribed to the change in drift intensity. Therefore, for the upper single null divertor configuration, a lower/higher plasma currents is more favorable for increasing and improving LH wave coupling in anti-clockwise/clockwise . The above studies are of great significance for optimizing and improving LH wave coupling in the extension of the plasma operation regime.

Predictive modeling of a lithium vapor box divertor in NSTX-U using SOLPS-ITER

Abstract The unmitigated heat flux in attached operation of a fusion power plant is predicted to be destructive to any solid divertor surface. Detachment, whereby the plasma pressure drops significantly before reaching the divertor target thus greatly reducing the heat flux and sputtering, will be necessary to ensure adequate lifetime of plasma facing components (PFCs). The lithium vapor box divertor aims to detach the divertor plasma via evaporating and condensing lithium surfaces. By evaporating lithium near or at the divertor plate and condensing it closer to the main chamber, a lithium vapor density gradient can be created. This density gradient ties energy losses to poloidal distance between the target and the detachment point. The radiation zone is then prevented from reaching the X-point as the lithium ionization rate decreases when the detachment front moves away from the divertor target. Here we present Scrape Off Layer Plasma Simulator (SOLPS) simulations of a lithium vapor box divertor using an NSTX-U equilibrium and PFC geometry. The parameters for the core boundary conditions, gas puff intensity, and heat and particle transport coefficients are chosen to match experimental values. Acceptable agreement with experimental Scrape-Off Layer (SOL) widths is found, indicating a reasonable choice of transport coefficients. In predictive simulations, lithium is added via evaporation at the target. Predictions for peak heat fluxes and upstream impurity concentration are given for a variety of evaporation rates. Target electron temperature is predicted to be able to be reduced to recombination levels (below 1 eV) for lithium evaporation rates of 1 ⋅ 1 0 23 Li/s, indicating detachment. Peak heat flux at the lower outer target could be reduced by as much as a factor of six while maintaining upstream lithium fractions below 2%. The prevention of lithium from reaching the midplane is shown to be due to an increase in frictional forces acting on the lithium from a deuterium gas puff. Lithium is also shown to be redeposited close to the evaporator which is favorable for initial tests and future capillary porous systems.

Ion cyclotron resonance heating fast and slow wave excitation and power deposition in edge plasmas with application to ITER

The antenna power coupling to the plasma centre and its possible deleterious coupling to the plasma edge are key parameters in an ion cyclotron resonance heating system. The influence on these parameters by the confluence between the slow and the fast magnetosonic waves is studied for the case of large machines. Until now, the modelling of the scrape off layer region has been calculated by ANTITER II, which contains only the fast wave description and where the confluence with the slow wave (S wave) is approximated by the Alfvén resonance. In the present study, a complete modelling of the slow and fast waves is made by ANTITER IV. The modelling by the two codes is compared and shows the important role of the Alfvén and the lower hybrid resonances for the excitation of large fields and associated power deposition at the edge of the plasma even far from the antenna location. The ANTITER IV modelling is thereafter applied to the case of the ITER antenna with a reference density profile and heating parameters. A comparative study is made for the edge power deposition and the excitation of large fields for different toroidal phasing cases of the antenna. This study also takes into account the tilting of the antenna array with respect to the total magnetic field in front of the antenna. If the Faraday screen is field-aligned, the excitation of the S wave occurs at the wave confluence; however, in the case of non-alignment the antenna directly excites the S wave. This effect is studied and quantified. All edge effects, even the direct excitation of S waves, can be strongly reduced by tailoring the current distribution in the straps of the antenna array. Resulting cases for the minimisation of edge power deposition in ITER and the reactor are studied.

Spatially dependent modeling and simulation of runaway electron mitigation in DIII-D

New simulations with the Kinetic Orbit Runaway electron (RE) Code KORC show RE deconfinement losses to the wall during plasma scrape off are the primary current dissipation mechanism in DIII-D experiments with high-Z impurity injection, and not collisional slowing down. The majority of simulations also exhibit an increase in the RE beam energy due to acceleration by the induced toroidal electric field, even while the RE beam current is decreasing. In this study, KORC integrates RE orbits using the relativistic guiding center equations of motion, and incorporates time-sequenced, experimental reconstructions of the magnetic and electric fields and line integrated electron density to construct spatiotemporal models of electron and partially-ionized impurity transport in the companion plasma. Comparisons of experimental current evolution and KORC results demonstrate the importance of including Coulomb collisions with partially-ionized impurity physics, initial RE energy, pitch angle, and spatial distributions, and spatiotemporal electron and partially-ionized impurity transport. This research provides an initial quantification of the efficacy of RE mitigation via injected impurities, and identification of the critical role played by loss of confinement due to plasma scrape off on the inner wall as compared to the relatively slow collisional damping.

Integrated modelling: Coupling of surface evolution and plasma-impurity transport

During the interaction of the scrape off layer (SOL) plasma with the first wall the evolution of both wall and plasma are tightly coupled: The erosion of the first wall leads to an impurity concentration in the plasma which affects the particle and power balance in the plasma. In turn the impurities, when leaving the plasma via transport, can form deposits and mixed materials, far away from their initial source location. These deposits can be eroded, allowing the impurities to stepwise migrate through the fusion device until they end up at a location where the plasma at the wall is cold enough and no further erosion occurs. To describe these processes an integrated model of surface evolution and plasma transport of impurities is needed. The WallDYN code achieves this required coupling of processes by parameterising the output of surface evolution- and plasma impurity-migration-codes by analytical models. For a given fixed background plasma it evolves the surface composition, derives impurity flux into and from the plasma and can from this derive the impurity densities in the plasma. This paper will show the importance of including this recycling of impurities at the wall in impurity migration modelling: The 13 CH 4 seeding experiment performed in the Wendelstein 7-X Stellerator is modelled using the recent extension of the WallDYN code to 3D plasma and wall geometries. A comparison with post mortem analysis of the 13 C deposition shows both qualitative and quantitative agreement with the WallDYN calculations.

Low-n global ideal MHD instabilities in the CFETR baseline scenario

This article reports an evaluation on the linear ideal magnetohydrodynamic (MHD) stability of the China Fusion Engineering Test Reactor (CFETR) baseline scenario for various first-wall locations. The initial-value code NIMROD and eigen-value code AEGIS are employed in this analysis. Despite the distinctly different approaches in modeling the scrape off layer region, the dominant growth in each of the low-n (n = 1 − 10) modes are consistent between the two codes. The higher-n modes are dominated by ballooning modes and localized in the pedestal region, while the lower-n modes have more prominent external kink components and broader mode profiles. The influence of the plasma-vacuum profile and wall shape are also examined using NIMROD. In the presence of a resistive wall, the low-n ideal MHD instabilities are further studied using AEGIS. For the designed first-wall location, the n = 1 resistive wall mode is found to be unstable, which could be fully stabilized by uniform toroidal rotation above the 2.9% core Alfvén speed.

Ion temperature measurement techniques using fast sweeping retarding field analyzer (RFA) in strongly intermittent ASDEX Upgrade tokamak plasmas.

This manuscript presents a new method of interpreting the ion temperature (Ti) measurement with a retarding field analyzer (RFA) that accounts for the intermittent/turbulent nature of the scrape off layer (SOL) plasmas in tokamaks. Fast measurements and statistical methods are desirable for an adequate description of random fluctuations caused by such intermittent events as edge localized modes (ELMs) and blobs. We use a RFA that can sweep its current-voltage (I-V) characteristics with up to 10 kHz. The RFA uses an electronics compensation stage to subtract the capacitive pickup due to the finite connecting cable capacitance, which greatly improves the signal-to-noise ratio. In the 10 kHz case, a single I-V characteristic is obtained in time, which is an order of magnitude faster than the ELM cycle. The fast sweeping frequency allows us to reconstruct the Ti probability density function (PDF), which we use as the Ti representation. The boundary conditions that we place on the I-V characteristics when calculating the Ti values impact the resulting Ti PDF. If the boundaries are insensitive to the plasma fluctuations, then the most probable Ti value of the PDF (20 eV-25 eV) is similar to the Ti value obtained via the classical conditional averaging method (20 eV-27 eV). However, if the boundary conditions follow the fluctuations, then the PDF-based method gives a substantially higher most probable Ti value (35 eV-60 eV). Overall, we show that a fast sweeping RFA diagnostic should be used in intermittent SOL plasmas to reconstruct the PDF for accurate Ti measurements.