The impact of ELMs on the ITER divertor

Leonard, A.W.; Herrmann, A.; Itami, K.; Lingertat, J.; Loarte, A.; Osborne, T. H.; Suttrop, W.

doi:10.1016/s0022-3115(98)00522-4

Cited by 137 publications

(131 citation statements)

References 22 publications

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“…The most likely reason for this is that type I ELMs have larger footprints than type III ELMs at the divertor [25] and therefore if the ELM moves past the divertor RFEA then signal will be seen for a longer time. As has already been shown, each type I ELM measured has a different current structure measurement which appears to be several filaments moving past the probe radially.…”

Section: Fst Type I Elm Ti Measurementsmentioning

confidence: 99%

Scrape-off layer ion temperature measurements at the divertor target during type III ELMs in MAST measured by RFEA

Elmore

Allan

Fishpool

et al. 2015

Journal of Nuclear Materials

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In future nuclear fusion reactors high heat load events, such as edge-localised modes (ELMs), can potentially damage divertor materials and release impurities into the main plasma, limiting plasma performance. The most difficult to handle are type I ELMs since they carry the largest fraction of energy from the plasma and therefore deposit the largest heat flux at the target and on first wall materials. Knowing the temperature of the ions released from ELM events is important since it determines the potential sputtering they would cause from plasma facing materials. To make measurements of T i by retarding field energy analyser (RFEA) during type I ELMs a new operational technique has been used to allow faster measurements to be made; this is called the fast swept technique (FST). The FST method allows measurements to be made within the time of the ELM event which has previously not been possible with T i measurements. This new technique has been validated by comparing it with a slower average measurement previously used to make ion temperature measurements of ELMs. Presented here are the first T i measurements during Type I ELMs made at a tokamak divertor. Temperatures as high as 20 eV are measured more than 15 cm from the peak heat flux of an ELM, in a region where no inter-ELM current is measured by the RFEA; showing that ELM events cause hot ions to reach the divertor target far into the scrape off layer. Fast camera imaging has been used to investigate the type of ELM filaments that have been measured by the divertor RFEA. It is postulated that most of the ion temperatures measured in type I ELMs are from secondary ELM filaments which have not been previously identified in MAST plasmas.2

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Section: Fst Type I Elm Ti Measurementsmentioning

confidence: 99%

Scrape-off layer ion temperature measurements at the divertor target during type III ELMs in MAST measured by RFEA

Elmore

Allan

Fishpool

et al. 2015

Journal of Nuclear Materials

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“…In type I edge localize modes (ELMS), a sudden collapse of the pressure gradient caused by rmicroinstability-driven turbulence leads to strong power loss across the separatrix of 2 6% of the total plasma stored energy with up to 80% being deposited on the divertor PFCs between 0.1-1 ms [16]. Plasma disruptions are another important transient event.…”

Section: Erosionmentioning

confidence: 99%

An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices

Hartwig

Barnard

Lanza

et al. 2013

Review of Scientific Instruments

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Plasma-material interactions (PMI) in magnetic fusion devices such as fuel retention, material erosion and redeposition, and material mixing present significant scientific and engineering challenges, particularly for the next generation of devices that will move towards reactor-relevant conditions. Achieving an integrated understanding of PMI, however, is severely hindered by a dearth of in-situ diagnosis of the plasma-facing component (PFC) surfaces. To address this critical need, this thesis presents an accelerator-based diagnostic that nondestructively measures the evolution of PFC surfaces in-situ. The diagnostic aims to remotely generate isotopic concentration maps that cover a large fraction of the PFC surfaces on a plasma shot-to-shot timescale.The diagnostic uses a compact, high-current radio-frequency quadrupole accelerator to inject 0.9 MeV deuterons into the Alcator C-Mod tokamak. The tokamak magnetic fields in between plasma shots are used to steer the deuterons to PFCs where the deuterons cause high-Q nuclear reactions with low-Z isotopes ~5 pim into the material. Scintillation detectors measure the induced neutrons and gammas; energy spectra analysis provides quantitative reconstruction of surface concentrations. An overview of the diagnostic technique, known as accelerator-based in-situ materials surveillance (AIMS), and the first AIMS diagnostic on the Alcator C-Mod is given; a description of the complementary simulation tools is also provided. Experimental validation is shown to demonstrate the optimized beam injection into the tokamak, the quantification of PFC surfaces isotopes, and the measurement localization provided by magnetic beam steering. Finally, the first AIMS measurements of fusion fuel retention are presented, demonstrating the local erosion and codeposition of deuterium-saturated boron surface filns. The finding confirms that deuterium codeposition with boron is insufficient to account for the net fuel retention in Alcator C-Mod.

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“…A comparison of IR, LP and TC time-averaged heat flux profiles for a typical ELMy H-mode on JET is shown in Fig.4.2.1-1. Profiles are usually time-averaged in ELMy H-mode (in part because few diagnostics have the necessary time resolution to distinguish ELMs) but observations from fast diagnostics indicate that ELMs either do not significantly modify the inter-ELM profile (outer divertor) [14,15,16] or dominate it (typical of the inner divertor) [2,8]. The outer target receives the majority of the divertor heat load, concentrated in a narrower profile than at the inner target (typical out:in power asymmetry ~ 2.5:1 and out:in peak heat flux asymmetry ~ 5:1) [17].…”

Section: Steady State Divertor Power Deposition Profilesmentioning

confidence: 99%

“…Initial studies of ELM energy losses [15] concluded that, in order to compare ELM size across devices and configurations and to project to future tokamaks, the energy lost at an ELM should be normalised to an estimate of the bulk plasma energy which is potentially accessible to expulsion by ELMs. This is usually characterised by the pedestal plasma energy, which is the total pedestal pressure, ions plus electrons, times the plasma volume [15].…”

Section: Mechanism Of Type I Elm Energy Transport To Pfcsmentioning

confidence: 99%

“…Heat fluxes are typically derived using data from infra-red cameras (IR), Langmuir probe arrays (LP) and thermocouples (TC), all of which have interpretation issues (e.g. IR thermography is subject to the effect of surface layers and localised hot-spots [7, 8; 9, 10 [14,15,16] or dominate it (typical of the inner divertor) [2,8]. The outer target receives the majority of the divertor heat load, concentrated in a narrower profile than at the inner target (typical out:in power asymmetry ~ 2.5:1 and out:in peak heat flux asymmetry ~ 5:1) [17].…”

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confidence: 99%

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Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

282

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Abstract.Progress, since the ITER Physics Basis publication, in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are : energy transport in the SOL in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main chamber material elements, ELM energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low and high Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas : refinement of the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a 2 major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed. Introduction.This chapter outlines the significant progress achieved since the ITER Physics Basis in understanding basic scrape-off layer (SOL) and divertor processes in a tokamak. The interaction of plasma with first-wall surfaces will have considerable impact on the performance of fusion plasmas, the lifetime of plasma facing components, and the retention of tritium in next step Burning Plasma E...

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The impact of ELMs on the ITER divertor

Abstract: Portions of this document may be illegible electronic image products. Images are produced from the best available original document.

Cited by 137 publications

References 22 publications

Scrape-off layer ion temperature measurements at the divertor target during type III ELMs in MAST measured by RFEA

Scrape-off layer ion temperature measurements at the divertor target during type III ELMs in MAST measured by RFEA

An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices

Chapter 4: Power and particle control

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