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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Impact of toroidal magnetic field direction on integrated ELM-stable operation and divertor power exhaust via boron powder injection in EAST

We report the first in-depth comparison of the impact of toroidal magnetic field direction on solid boron injection used for Edge-Localized Mode (ELM) control, power exhaust, and core high-Z impurity control in the Experimental Advanced Superconducting Tokamak. With favorable ion ∇B drift towards the upper X-point in an upper-single-null configuration, boron injection effectively suppresses ELMs, produces a detachment of the inner divertor target, and leads to improved energy confinement. ELM suppression in this configuration is accompanied by the excitation of an Edge Harmonic Mode. In contrast, with unfavorable ion ∇B drift away from the upper X-point, boron injection also suppresses ELMs but leads to a more symmetric detachment state of both the inner and outer divertor targets, while plasma energy confinement is slightly degraded despite similar boron injection levels; a different low-frequency coherent mode without multiple harmonics is observed. Measurements from toroidally separated views show that the divertor response to boron injection is essentially toroidally symmetric, supporting the use of two-dimensional SOLPS-ITER modeling with a toroidally uniform impurity source. These experimental observations are qualitatively consistent with SOLPS-ITER simulations, which highlight the critical role of E × B drift effects in setting the Bt-dependent in–out asymmetry of detachment and in asymmetrically transporting particles and injected impurities within the scrape-off layer and private-flux region. These findings underscore the importance of drift physics and real-time wall conditioning in controlling low-Z impurity transport and optimizing edge solutions for integrated, ELM-stable, high-performance tokamak operation.

E × B drift↗
DOE OSTI2026

Variations in lithium vapor cave performance predictions due to radial transport and recycling assumptions

The lithium vapor cave is a detached divertor design that uses a single private flux region baffle to contain a dense cloud of lithium vapor to dissipate heat flux. Plasma flows are created via fuel gas puffing in order to minimize lithium contamination of the main plasma. Significant modeling using the 2D edge code SOLPS-ITER has already been performed, predicting that sufficient target heat flux reductions (q$^{max}_{Target}$ MW m −2 ) with ( $n$ Li /$n$ e ) LCFS < 0.05 is possible in a case with 90 MW m −2 unmitigated heat flux in NSTX-U. Low heat flux and low upstream concentration was found with a variety of combinations of divertor geometries, target recycling coefficients, upstream plasma parameters, lithium evaporation locations and deuterium fueling locations, with variations in performance found for each design choice. However, the most universal uncertainty of SOLPS-ITER simulations has until now remained unaddressed systematically, namely the cross-field anomalous particle and heat diffusivities. This article aims to bound the uncertainty in lithium concentration prediction as a result of the assumed deuterium cross-field transport and recycling. For the simulations presented here, a factor of 2.1 increase in the upstream lithium density prediction across a factor of four decrease to the assumed deuterium radial particle diffusivity is found. This result is compared across different assumed deuterium recycling coefficients, known to reduce with lithium injection. Upstream lithium density is found to vary by a factor of 2.4 across a feasible range of recycling coefficients at similar $n$$^{OMP,sep}_{e}$.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Radiative divertor detachment and impurity transport with nitrogen and neon seeding in KSTAR H-mode plasmas

Achieving core-edge compatible divertor detachment is a critical requirement for stable operation in future fusion devices. This study compares nitrogen and neon seeding in KSTAR H-mode plasmas with carbon walls, combining experiments and SOLPS-ITER modelling to evaluate their radiative dissipation and core-edge compatibility. Experimentally, N seeding achieved stronger divertor detachment, with larger reductions in target particle and heat fluxes, a higher divertor radiation fraction, and a lower core radiation fraction than Ne. In contrast, Ne seeding triggered a significant rise in core radiation followed by H–L back transitions, limiting the maximum total radiated power fraction to roughly half that of N. SOLPS-ITER simulations reproduced the experimental trends and revealed that the better core-edge compatibility of N arises from its higher divertor retention in addition to its higher cooling factor. The relative positions of the stagnation points of impurity poloidal velocity and ionization sources did not explain the different divertor compression. Instead, in the present modelling, the higher impurity parallel particle flux, resulting from the higher parallel impurity velocity, explains the stronger nitrogen impurity compression in the divertor region. The parallel temperature distribution with N was more favourable for achieving higher impurity parallel velocity than with Ne, because the impurity velocity is governed by modifications of the main ion flow due to friction and thermal forces, both of which strongly depend on the temperature. Ultimately, this behaviour is attributed to the strongly divertor-localized radiation of N. These results demonstrate that N is more effective than Ne in achieving radiative divertor detachment while maintaining low core contamination in KSTAR, consistent with observations in other present tokamaks.

KSTAR↗
DOE OSTI2022

Numerical Analysis of Liquid Metal MHD Flow and Heat Transfer for Open-Surface Li Divertor in FNSF

Within the ongoing U.S.-based program on the development of liquid metal plasma-facing components, numerical simulations and analyses are performed to address the feasibility of the open-surface Li divertor. In the previous scoping studies (Smolentsev, 2021), heat-removal capabilities of the divertor were assessed using a simplified flow model for a slug-type velocity profile and constant flow thickness. Here, new analyses take into account forces acting on the flowing Li layer. Three reduced-order mathematical models are applied under the conditions of the U.S. Fusion Nuclear Science Facility (FNSF) to access magnetohydrodynamic (MHD) flow development effects, velocity distribution, and surface waves: 1) fully developed MHD flow; 2) quasi-2-D developing MHD flow; and 3) multiphase MHD flow. The obtained results for MHD flows and the surface heat flux computed with the plasma code scrape-off layer plasma simulation for ITER (SOLPS-ITER) are then used as input data to compute the temperature distribution in the divertor by solving the convection–diffusion energy equation.

Smolentsev, Sergey↗
DOE OSTI2020

Modeling ExB drift transport in conceptual slot divertor configurations

At DIII-D, a slot divertor concept, called small-angle-slot (SAS), is under development, aiming to enable detachment at relatively low plasma edge density. We report on simulations using the SOLPS-ITER two-dimensional edge code to examine the performance of conceptual “SAS 2” slot configurations. The focus of the analysis is on E × B drift effects on upstream density at detachment (UDD), with detachment marked by electron temperature T e ≤ 3 eV at the outer strike point (OSP). With toroidal field such that radial E × B drift carries particles from the OSP towards the private flux region (PFR), placing the OSP near the inner slot wall gives ≈20% lower UDD than having the OSP near the outer wall. The inner wall effectively traps the radial E × B drift flux, resulting in low T e and associated radial electric field in the PFR, and thus small losses from the slot to the inner target via poloidal E × B drift flux. With toroidal field reversed such that radial E × B drift is reversed, OSP placement near the inner wall gives ≈10% lower UDD than OSP placement near the outer wall. Although radial E × B flux is from the OSP towards the outer wall, this flux largely escapes the slot, raising the UDD. Finally, a change in the slot shaping is suggested with the goal of eliminating such E × B-driven particle losses from the slot.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

Role of E × B Drift in Divertor Detachment Control via Boron Powder Injection on EAST

Here, the effects of B powder injection on plasma detachment about EAST discharge were studied by using SOLPS-ITER code package with the effects of E × B drifts considered. The simulation results show that plasma detachment occurs at the inner target in favourable toroidal magnetic field (B t ) direction at a relatively low B powder flow rate, one order of magnitude lower than that at the outer target. In a similar scenario with unfavourable B t , it is found that the detachment thresholds of B flow rate for both the inner and outer targets are close and of the same order as that for the outer target with favourable Bt. In favourable B t direction at B powder flow rate of 1.2 × 10 21 atoms/s, a localized, broadened high-density region is formed near the inner target benefitted by the injection location and the E × B drift, and a radiation-intensified zone, mostly contributed by B 1+ and B 2+ , occurs there. The E × B drift facilitates plasma detachment at the inner target and simultaneously amplifies the in–out divertor asymmetry. In addition, the simulation results with three different injection locations show that the injection from outer strike point leads to the lowest Z eff inside the separatrix and has an intermediate flow rate for detachment at the outer target, comparing with the X-point and upstream locations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Optical diagnostic design for measuring the radiation front with a mid-leg pumped divertor on DIII-D

The design of an optical diagnostic system to localize the radiation front in a mid-leg pumped divertor configuration on DIII-D is presented. Divertor detachment is a key mechanism for handling power exhaust in tokamaks, and mid-leg pumping offers a promising approach to radiatively dissipate power while maintaining acceptable core performance. To predict the location of the radiation front and design a spectroscopic diagnostic to measure its position, a database of SOLPS-ITER simulations across a range of input powers and gas puffing rates representative of DIII-D operation was employed. These simulations provide self-consistent plasma backgrounds for a Cherab-Raysect synthetic diagnostic framework, which incorporates detailed tokamak geometry and physically accurate ray-tracing. Synthetic simulations of optical sightlines and viewing cone geometry were used to calculate line-integrated emission from the plasma, particularly of the C III 465 nm line, to serve as a proxy for the T e ≈ 7 − 10 eV temperature region associated with the onset and evolution of divertor detachment. Neutral deuterium emission is also evaluated for comparison. The synthetic diagnostic is used to assess the performance of existing DIII-D optical systems, including filterscopes and the Multichord Divertor Spectrometer, and to optimize line-of-sight placement within mechanical and installation constraints. The results provide quantitative guidance for diagnostic implementation on DIII-D and demonstrate the advantage of integrated synthetic diagnostics for divertor design studies and future advanced divertor concepts.

Cherab↗
DOE OSTI2021

Numerical simulation study of impurity B transport during real-time B powder injection in EAST

To understand the beneficial effects of impurity boron (B) on the plasma, the transport of B species in both atomic and ionic states in the edge plasma and the ensuing evolution of the background plasma are simulated during real-time B powder injection by using two-dimensional fluid code SOLPS-ITER. The initial background plasma profiles for the simulation are reconstructed based on an EAST upper-single null discharge (shot No. 70601). With the injection rate of B atoms, 4.0 × 10 21 atoms/s, the simulation shows that effective charge number is less than 1.1 at the inner boundary R - R sep = -5.2 cm (background plasma density, 5.1 × 10 19 m -3 ) at the outer mid-plane indicating that the B impurity does not have much influence on the energy confinement. The present work finds that the impurity B exhibits very specific features in the edge plasma, mostly present in the narrow region just outside the separatrix. The particular distribution pattern can reduce the heat flux greatly, leading to the divertor detachment. The divertor detachment can be achieved easily under an injection rate of B powder in a reasonable range. This study suggests that the impurity B has a promising potential application in future tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Simulations of divertor designs that spatially separate power and particle exhaust using mid-leg divertor particle pumping

Predictive design modeling of a Dissipation-Focused Divertor for future operation in DIII-D reveals that increasing the poloidal distance of the pump duct entrance from the target surface along the low-field side divertor baffle increases neutral compression and modifies the spatial distribution of power dissipation. With a divertor pump located mid-leg between the target and the X-point, SOLPS-ITER boundary plasma simulations without drifts predict the formation of a dense neutral cloud near the target with > 30x higher neutral compression in detachment, a more stable detachment front located further from the target, and ~25% lower outer midplane separatrix density required for detachment onset, compared to a pump located in the scrape-off layer at the target surface. Up to 19 MW of power flowing into the divertors is modeled using the following two numerical implementations for particle pumping: a specified fraction of particles incident on variable wall sections of the plasma grid is removed from the computational domain (so-called albedo pumping), and a pump duct is modeled which includes dynamics of kinetic neutrals in the duct. The simulations show that the detachment front is located between the divertor target and the X-point and is relatively stable near the pump entrance, without a strong dependence on gas puff rate or injected power. The mid-leg pump design spatially separates the two primary functions of a divertor (power handling and particle exhaust), with the majority of power dissipation occurring near the target plate and particle exhaust taking place further upstream. The benefit of enhanced dissipation using mid-leg pumping comes at the cost of a higher outer midplane separatrix density for a given amount of particle injection.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Design and physics basis for the upcoming DIII-D SAS-VW campaign to quantify tungsten leakage and transport in a new slot divertor geometry

A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot divertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) divertor is being coated with 10–15 μ m of tungsten. New spectroscopic viewing chords with in-vacuo optics will measure the W gross erosion source from the divertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS divertor is changed from a flat to a ‘V’ shape. New SOLPS-ITER/DIVIMP simulations conducted with drifts using the planned ‘V’ shape predict a substantial reduction in W sourcing and SOL accumulation in either B × ∇B direction relative to either the old SAS divertor shape or the open, lower divertor. Dedicated studies are planned to carefully characterize the level of W sourcing, leakage, and scrape-off-layer (SOL) accumulation in DIII-D over a wide range of plasma scenarios. Various actuators will be assessed for their efficacy in further reducing high-Z impurity sources and leakage from the slot divertor geometry. Finally, this coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z divertor geometries for next-step devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

First evidence of dominant influence of E × B drifts on plasma cooling in an advanced slot divertor for tokamak power exhaust

Addressing power exhaust in tokamaks is presently recognized as one of the major remaining open issues for the development of fusion reactors. At the forefront of this endeavor is the effort to develop an advanced divertor by maximizing dissipation of plasma power and momentum inside the divertor. Here, we demonstrate, for the first time, that the electromagnetic (E × B) drifts exert a key influence on plasma and gas dynamics in a new advanced slot divertor in the DIII-D tokamak, named the small angle slot (SAS). SAS leverages the effect of drifts to achieve a highly dissipative divertor with electron temperature T e ≲ 10 eV over a wide range of plasma densities, for ion B × ∇B away from the divertor, as used for advanced tokamak operation on DIII-D. Modeling with the SOLPS-ITER code shows that for this drift direction, the E × B flow carries particles toward the outer common flux region (CFR) via the private flux region (PFR), reinforcing neutral recycling and enhancing divertor dissipation. In contrast, for the opposite field direction, the E × B flow carries particles away from the outer CFR into the PFR, offsetting the anticipated SAS geometric effects. Furthermore, this finding is an important step in the understanding of the behavior of advanced divertors for achieving a power exhaust solution for fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Effects of radial transport on divertor power and particle flux widths under different operational regimes in EAST

Here, a study of the effects of radial transport on the outer divertor particle and power flux widths (λ js and λ q ) is carried out using plasmas under different operational regimes in the Experimental Advanced Superconducting Tokamak (EAST). In the EAST experiments, the λ js values measured during the grassy intra- and inter-ELM phases are similar, and the averaged value of λ js during grassy ELMy discharges is found to be smaller than that of the intra-ELM phase of type-I ELMy discharges but larger than that of the inter-ELM phase of type-I ELMy discharges. Simulations of scans of the radial particle and heat transport coefficients, D and χ e , performed using the BOUT++ transport code show that both λq and λjs increase with D and χe, especially when D or χ e is larger than a threshold value, indicating that background turbulence starts to matter in the determination of λ q and λ js when D or χ e surpasses the threshold value. A comparison between the simulation of the D scan and the experimental results shows that the different values of λ js obtained under different plasma operational regimes are probably due to the different intensities of background turbulence, which is beyond the scope of Goldston's heuristic drift-based model as reported by Goldston et al (2012 Nucl. Fusion 52 013009), since their model was derived by assuming that turbulent transport does not exceed a certain level. The ratio of λ js to λ q is greatly affected by the radial transport; however, simulations carried out using Scrape-off Layer Plasma Simulation (SOLPS) show that divertor geometry and plasma density have big influences on λ js , and thus could also affect the ratio of λ js to λ q .

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Numerical assessment of the new V-shape small-angle slot divertor on DIII-D

The small-angle slot (SAS) divertor of the DIII-D tokamak, and its upcoming upgrade, the V-shape small-angle slot (SAS-V) divertor, are numerically investigated using the SOLPS-ITER code package, including the effect of particle drifts, for a range of plasma density, heating power, strike point position in the slot, and for both magnetic field directions. The simulations show that the electron temperature near the strike point is reduced in SAS-V compared to SAS, for both magnetic field directions, such that SAS-V achieves divertor detachment at a lower value of the outboard mid-plane separatrix electron density. The detachment threshold is lower because the V-shape focuses recycling neutrals on the V-end, densifying and cooling the plasma in the slot. At sufficiently high density, the V-shape also reduces the radial gradient of the temperature profile at the target, which in turns reduces the radial electric field and the E x B drift velocities, further densifying and cooling the plasma in the slot and leading to detachment. The V-shape effect, however, is reduced for higher heating power. With more heating power, the detachment density increases, reducing the ionization mean free path of recycled neutrals, which therefore become less sensitive to target shape changes. Furthermore, this suggests that in a fusion reactor, where the heating power is high, optimization of the divertor target shape needs to be combined with other strategies to lower the detachment density, such as in-slot injection of low-Z impurities.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Impurity leakage and radiative cooling in the first nitrogen and neon seeding study in the closed DIII-D SAS configuration

A comparative study of nitrogen versus neon has been carried out to analyze the impact of the two radiative species on power dissipation, SOL impurity distribution, divertor and pedestal characteristics. The experimental results show that N remains compressed in the divertor, thereby providing high radiative losses without affecting the pedestal profiles and displacing carbon as dominant radiator. Neon, instead, radiates more upstream than N thus reducing the power flux through the separatrix leading to a reduced ELM frequency and compression in the divertor. A significant amount of neon is measured in the plasma core leading to a steeper density gradient. The different behavior between the two impurities is confirmed by SOLPS-ITER modeling which for the first time at DIII-D includes multiple impurity species and a treatment of full drifts, currents and neutral–neutral collisions. The impurity transport in the SOL is studied in terms of the parallel momentum balance showing that N is mostly retained in the divertor whereas Ne leaks out consistent with its higher ionization potential and longer mean free path. This is also in agreement with the enrichment factor calculations which indicate lower divertor enrichment for neon. The strong ionization source characterizing the SAS divertor causes a reversal of the main ions and impurity flows. The flow reversal together with plasma drifts and the effect of the thermal force contribute significantly in the shift of the impurity stagnation point affecting impurity leakage. This work provides a demonstration of the impurity leakage mechanism in a closed divertor structure and the consequent impact on pedestal. Since carbon is an intrinsic radiator at DIII-D, in this paper we have also demonstrated the different role of carbon in the N vs Ne seeded cases both in the experiments and in the numerical modeling. Here, carbon contributes more when neon seeding is injected compared to when nitrogen is used. Finally, the results highlight the importance of accompanying experimental studies with numerical modeling of plasma flows, drifts and ionization profile to determine the details of the SOL impurity transport as the latter may vary with changes in divertor regime and geometry. In the cases presented here, plasma drifts and flow reversal caused by high level of closure in the slot upper divertor at DIII-D play an important role in the underlined mechanism.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Predicting tungsten erosion and leakage properties for the new V-shaped small angle slot divertor in DIII-D

Impurity transport modeling of the new tungsten (W)-coated, V-shaped small angle slot (SAS) divertor in the DIII-D tokamak was conducted using the SOLPS-ITER plasma edge code package and the DIVIMP impurity tracking code. The inboard baffle of the current SAS divertor will be shifted closer to the outboard baffle, creating a V-corner at the slot vertex. In addition, the outboard baffle will be coated with 10–15 μm of W for experiments studying high-Z sourcing and leakage in a closed divertor. Modeling of the 'SAS-VW' divertor predicts that these changes to the inner baffle will reduce W gross erosion by 40× relative to the existing SAS divertor when the outer strike point (OSP) is at the V-corner and the ion B × ∇B drift is towards the divertor, driven primarily by significant cooling near the slot vertex. Most W erosion in SAS-VW is expected to occur near the slot entrance, which may pose a higher risk to core contamination than W eroded deeper in the slot. Furthermore, adding a new sheath-based prompt redeposition model increases the sensitivity of redeposition estimates to near-target plasma conditions and may provide more accurate predictions of net erosion. Moving the OSP outboard from the slot vertex ~4 cm onto the W-coated region yields a 40× increase in the gross erosion rate and a 50% decrease in the core leakage fraction. Thus slight variations in strike point location may counteract the potential benefits of the tightly-baffled V slot on minimizing erosion. This impurity transport modeling provides useful guidance for future experiments on the SAS-VW divertor focused on high-Z erosion/redeposition, scrape-off layer transport, and core leakage.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Increased radiation due to non-coronal effects on DIII-D and MAST-U with varying input power

Through SOLPS-ITER simulations of DIII-D and MAST-U, an X-divertor (XD) on DIII-D and a super X-divertor (SXD) on MAST-U were shown to have increased carbon emissivity (P Rad /n e n I ) over corresponding standard divertors (SD) at similar degrees of partial detachment. The reasons behind the increased emissivity in the DIII-D XD and SXD are analyzed using a simple 0D transport model. From the transport model, it is seen that a major cause of the increased emissivity in the XD and SXD over the SDs is a shorter impurity confinement time. An additional cause (for the SXD) is an increase in the ratio of neutral hydrogen to electron density. The input power (P in ) was varied and the XD had a higher emissivity at the higher P in , unlike the SDs which had the emissivity decrease with increasing P in . A basic geometrical reason is given to explain both the benefits of the XD over the SD as well as the increase in the XD's emissivity with P in .

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Energy and particle balance during plasma detachment in a long-leg divertor configuration

Abstract Comprehensive studies of energy and particle balances in the transition to plasma detachment in an alternative divertor configuration with long outer legs are shown. Numerical simulations are performed with the 2D code suite SOLPS 4.3, using a disconnected double null grid with narrow, tightly baffled long poloidal leg divertors at the outer lower target and outer upper target. A particle count scan is performed using the ‘closed gas box’ model, where the tunable parameter in the simulations is the total number of deuterium particles in the simulation space and all other parameters are held fixed, including a constant input power and trace neon impurity radiation, to assess the physics of the transition to detachment in the system as the particle count increases. Three main aspects of the physics of divertor detachment are addressed: the criteria for the local onset of divertor detachment in each of the divertors, the distribution of heat flux and other plasma parameters between the four divertors as each divertor transitions to detachment, and the role of perpendicular transport in the transition to the detached regime. A synergistic mechanism by which the cross-field transport is reduced by factors associated with the onset of plasma recombination effects is identified. These results are compared to the existing understanding of the physics of the transition to plasma detachment in standard divertors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Evaluation of ITER divertor shunts as a synthetic diagnostic for detachment control

Reliable diagnostics that measure the detached state of the ITER divertor plasma will be necessary to control heat flux to the divertor targets during steady state, burning plasma operation. This paper conducts an initial exploration into the feasibility of the divertor shunt diagnostic as a lightweight, robust, and real-time detachment sensor. This diagnostic is a set of shunt lead pairs that measure the voltage drop along the divertor cassette body, from which the plasma scrape-off layer (SOL) current is calculated. Using SOLPS-ITER simulations for control-relevant ITER plasma scenarios, the thermoelectric current magnitude along the SOL is shown to decrease significantly with the onset of partial detachment at the outer divertor target. Electromagnetic modelling of a simplified divertor cassette is used to develop a control-oriented inductance-resistance circuit model, from which SOL currents can be calculated from shunt pair voltage measurements. The sensitivity and frequency-response of the resulting system indicates that the diagnostic will accurately measure SOL thermoelectric currents during ITER operation. These currents will be a good measure of the detached state of the divertor plasma, making the divertor shunt diagnostic a potentially extremely valuable and physically robust sensor for real-time detachment control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗