Effects of divertor geometry and chemical sputtering on impurity behaviour and plasma performance in JET

Guo, Heng; Matthews, G. F.; Coffey, I.; Erents, S.K.; Groth, M.; Harbour, P.J.; Hellermann, M. G. von; Hillis, D.L.; Hogan, J.; Horton, L.D.; Ingesson, L. C.; Lawson, K.; Lingertat, J.; Maggi, C. F.; McCracken, G.M.; Monk, R.D.; Morgan, P. D.; Stamp, M.; Stangeby, P.C.; Taroni, A.; Vlases, G.

doi:10.1088/0029-5515/40/3/308

Cited by 30 publications

(19 citation statements)

References 49 publications

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“…Integrated modelling of helium transport in the core, edge and pumping plenum using the COCONUT [245], B2-EIRENE [246,247,248], EDGE2D [217], and DIVIMP/NIMBUS [240,249] codes has shown qualitative agreement with experimental observations, which gives some confidence in the modelling predictions for ITER, although further work in this area to refine the predictions is on-going.…”

Section: Helium Exhaust and Noble Gas Impurity Enrichmentmentioning

confidence: 82%

“…Experiments in many divertor tokamaks have shown that increasing the divertor closure has led to larger divertor neutral pressures [211,212,179,62,213], as shown in Fig In general, the predictions with respect to the behaviour of deuterium neutral and recycling impurity exhaust by edge modelling codes have been confirmed by the experiments in tokamaks. In particular, the dependences of the neutral particle and impurity exhaust on the details of the divertor geometry and on the local magnetic flux surfaces have been confirmed [214,69,215,65,216,217]. The neutral pressure in the pumping plenum can be very sensitive to the exact location of the divertor strike point [69,215,216] when the neutral transport is kinetic (termed 'ballistic').…”

Section: Neutral Pressure Controlmentioning

confidence: 94%

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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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Section: Helium Exhaust and Noble Gas Impurity Enrichmentmentioning

confidence: 82%

Section: Neutral Pressure Controlmentioning

confidence: 94%

Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

282

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“…Implicit to this section, is the assumption that the wall components and conditioning are similar to the JET walls. For example, the common JET algorithm [7] for the impurity source is a carbon wall and divertor with a chemical sputtering yield of 0.5 of the Toronto value [8]. The factor of Ω was proposed in Ref.…”

Section: -Carbon Sputter Ingmentioning

confidence: 99%

“…The factor of Ω was proposed in Ref. [7] to account for the JET carbon light signals and may represent either a calibration factor in the JET instrumentation or the sputtering coefficients. This assumption, with the 22/01/0411:44 16 empirical screening, predicts JET carbon core contamination similar to the experimental values [2].…”

Section: -Carbon Sputter Ingmentioning

confidence: 99%

Diverted tokamak carbon screening: scaling with machine size and consequences for core contamination

et al. 2004

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Abstract:Plasma impurity content depends upon the impurity sources, fuelling efficiency, and confinement. In JET, carbon is the primary impurity, and its fuelling efficiency has been studied using methane gas injection and modeled with the SOL codes: DIVIMP and EDGE2D. In this paper, EDGE2D modeling of similar AUG experiments and projections to ITER are described. The parameters have been identified which govern the size scaling of carbon screening.Size scaling is complex. For carbon injected from the main chamber, the important factors include: the SOL temperature, the magnitude of the thermal force at the divertor entrance, and the parallel distance to the divertor. For carbon injected at the strike points, the intersection of the carbon ionization region with the region of strong thermal force determines the carbon fuelling efficiency ITER projects to have much better carbon screening than JET. The ITER SOL is hotter so that main chamber carbon is ionized further from the separatrix making the calculated carbon fuelling efficiency lower. Also, the carbon originating near the strike point has less chance of escaping the divertor by factors of about 100. The carbon sputtering is projected to be larger by similar factors, making the projected ITER core contamination similar to JET. However, that result is based upon the assumption that the wall materials have similar composition and behavior as observed on JET. A general result is that the core contamination at fixed total sputtering rate and core impurity confinement increases when the fraction of carbon ionized in the main chamber SOL increases, and decreases for larger machine size and higher density operation.

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“…The best explanation for the higher yield is that the chemical sputtering yield is increased due to the higher base temperature of the Mark II divertor target plate [34]. Specific experiments with lower wall temperature in Mark II support this explanation [35].…”

Section: Impurity Behaviourmentioning

confidence: 99%

Edge transport barrier in JET hot ion H modes

Guo¹,

Lomas²,

Parail³

et al. 2000

Nucl. Fusion

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Abstract. The effects of changing beam and plasma species on the edge transport barrier are investigated for ELM-free hot ion H mode discharges from the recent DT experiments on JET. The measured pressure at the top of the pedestal is higher for mixed deuterium and tritium and pure tritium plasmas over and above the level measured in pure deuterium plasmas at the same heating power. The pedestal pressure increases with beam tritium concentration for mixed deuterium-tritium beam injection into deuterium plasmas where the measured edge tritium concentration remains low. Alpha heating plays a significant role in the core of such plasmas, and the possible impact on the edge is discussed together with possible direct isotopic effects. Heuristic models for the transport barrier width are proposed, and used to explore a wider range of edge measurements including full power DD and DT pulses. This analysis supports the plasma current and mass dependence for a barrier width set by the orbit loss of either thermal or fast ions, though it does not unambiguously distinguish between them. The fast ion hypothesis could well account for some of the JET observations, though more theoretical work and direct experimental measurement would be required to confirm this. An ad hoc model for the power loss through the separatrix, P loss ∝ n 2 edge Z eff ,edge I −1 p , is proposed based on neoclassical theory, a ballooning limit to the edge gradient and a barrier width set by the poloidal ion gyroradius. Such a model is compared with experimental data from JET. In particular, the model ascribes the systematic difference in loss power between the Mark I and Mark II divertors to the change in the measured Z eff . This change in Z eff is consistent with the observed change in impurity production, which is described in some detail, together with a possible explanation provided by the temperature dependence of chemical sputtering.

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Effects of divertor geometry and chemical sputtering on impurity behaviour and plasma performance in JET

Cited by 30 publications

References 49 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Diverted tokamak carbon screening: scaling with machine size and consequences for core contamination

Edge transport barrier in JET hot ion H modes

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