Basic divertor operation in ITER-FEAT

Kukushkin, A. S.; Pacher, H.D.; Janeschitz, G.; Loarte, A.; Coster, D.; Matthews, G. F.; Reiter, D.; Schneider, R.; Zhogolev, V.E.

doi:10.1088/0029-5515/42/2/310

Cited by 61 publications

(70 citation statements)

References 9 publications

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“…A number of modelling results on the performance of the ITER divertor have been obtained in last five years [511,516,253,29,517,20,513,512,515] as new experimental results have been obtained, further physical effects have been included and a design optimisation of ITER toward lower cost has been performed. The primary aim of the studies has been the identification of the major control parameters, the exploration of the operational window, the investigation of the effect of divertor geometry, and the development of integrated modelling of the whole plasma by coupling of the edge and core models at their interface.…”

Section: Modelling Resultsmentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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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: Modelling Resultsmentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…However, recent results have shown that the plasma fluxes on the wall can be higher than expected from this simple exponential decay, due to turbulent transport across the field lines [16]. Present modelling [11,17] …”

Section: Iii-1) Parameters Of the Plasma In Contact With Plasma Facinmentioning

confidence: 89%

“…This allows to access to the detached divertor regime [10], which is the reference for ITER divertor operation [11]. In the detached divertor regime, dissipation of the power conducted/convected by the plasma by the high densities due to intense hydrogen recycling, impurities radiation, and partial recombination of the impinging particle flux allows to keep peak power fluxes on the divertor targets under ~ 10 MWm -2 , while maintaining adequate central plasma purity [11]. The corresponding ITER plasma parameters along the divertor target estimated from modelling are shown in Fig.…”

Section: Iii-1) Parameters Of the Plasma In Contact With Plasma Facinmentioning

confidence: 99%

“…The corresponding ITER plasma parameters along the divertor target estimated from modelling are shown in Fig. 2 [11,12]. eV, due to acceleration by the plasma sheath potential.…”

Section: Iii-1) Parameters Of the Plasma In Contact With Plasma Facinmentioning

confidence: 99%

“…• plasma conditions expected from plasma codes [11] • erosion yields dependent on energy, temperature and ion flux…”

Section: Present Estimate For Iter Inventorymentioning

confidence: 99%

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Plasma–wall interaction: Important ion induced surface processes and strategy of the EU Task Force

Roth¹,

Tsitrone²,

Loarte³

2007

Nuclear Instruments and Methods in Physics Research Section B:

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Abstract:In future thermo-nuclear fusion devices, such as ITER (International Thermonuclear Experimental Reactor), the interaction of the plasma with surrounding materials in the vacuum vessel constitutes one of the main remaining engineering problems. The choice of materials is a crucial point, which will determine issues such as the plasma facing components lifetime before refurbishment or the tritium inventory build up in the vessel, which should be limited for safety reasons. In order to tackle these issues, the European Task Force on Plasma wall interaction has been implemented in the frame of EFDA (European Fusion Agreement) in the fall 2002 with the aim "to provide ITER with information concerning lifetime-expectations of the divertor target plates and tritium inventory build-up rates in the foreseen starting configuration and to suggest improvements, including material changes, which could be implemented at an appropriate stage."The EU PWI TF brings together the efforts of 24 European associations in the following fields of investigation :• Material erosion and transport in tokamaks • Tritium inventory and removal • Transient heat loads on plasma facing components • Dust production and removal • Associated modelling and diagnostic development This paper will present the organisation of the EU PWI TF. It will provide examples for the multitude of surface processes in plasma-wall interaction and present the status of knowledge concerning material erosion and hydrogen retention for the choice of ITER materials (Beryllium, Carbon and Tungsten).

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Numerical Analysis of Divertor Plasma for Demo‐CREST

Ishida

Maeki

Hiwatari

et al. 2010

Contrib. Plasma Phys.

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The numerical analysis of the demonstration fusion reactor Demo-CREST has been carried out; this analysis focuses on impurity seeding. Several design activities for DEMO have been carried out; however, its detailed divertor plasma analysis remains to be carried out. Therefore, in this study, we discuss the possibility of neon puffing in Demo-CREST to decrease the power load to the divertor plate by using the B2-EIRENE code. It has been shown that the radiation power loss by neon increases with upstream plasma density and that the peak power load to the divertor plate comes close to the allowable level by using the preliminary divertor configuration.

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Basic divertor operation in ITER-FEAT

Cited by 61 publications

References 9 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Plasma–wall interaction: Important ion induced surface processes and strategy of the EU Task Force

Numerical Analysis of Divertor Plasma for Demo‐CREST

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