Assessment of erosion and tritium codeposition in ITER-FEAT

Federici, G.; Brooks, J.N.; Coster, D.; Janeschitz, G.; Kukuskhin, A; Loarte, A.; Pacher, H.D.; Stober, J.; Wu, C.H.

doi:10.1016/s0022-3115(00)00627-9

Cited by 131 publications

(97 citation statements)

References 19 publications

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“…The basic impurity generation mechanisms [274] are: 1) Sputtering of the main chamber walls by neutral particle bombardment. These sources can peak in the vicinity of gas or neutral beam injection ports [286]. Chemical sputtering of carbon walls is usually larger than physical sputtering [275].…”

Section: Intrinsic Impurity Sourcesmentioning

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: Intrinsic Impurity Sourcesmentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…For the divertor target strike point zones the example design consists of 28×20mm sized carbon fibre composite (CFC) mono blocks with 46mm thickness stacked in vertical columns with 0.5mm poloidal gap [15]. The target plate area of 55m 2 [16] will require ≈90000 mono blocks. From the tile dimensions one obtains a hidden side face area of 330% of the plasma exposed CFC surface.…”

Section: Extrapolation To Itermentioning

confidence: 99%

“…For the divertor outside the strike-point zones, the baffles and the divertor dome, the example design consists of 10×10×10mm tungsten rods with 0.5mm gaps [17]. With a total tungsten area of 140m 2 [16] there are ≈ 1200000 W-rods with a hidden side face area of 360% of the plasma exposed W-PFC surface. The main chamber wall armour example design consists of 42×47mm sized beryllium tiles with 10mm thickness and 2mm spacing bonded to a cooling structure [17].…”

Section: Extrapolation To Itermentioning

confidence: 99%

“…The main chamber wall armour example design consists of 42×47mm sized beryllium tiles with 10mm thickness and 2mm spacing bonded to a cooling structure [17]. With an ITER primary wall area of 690m 2 [16] the main chamber Be armour will have ≈300000 beryllium tiles. From the tile dimensions one derives a hidden side face area of 90% of the plasma exposed surface.…”

Section: Extrapolation To Itermentioning

confidence: 99%

“…Because in the present design carbon based materials will be used for the divertor target plates, eroded carbon atoms will migrate to other locations and co-deposit with the fuel isotopes. It is not clear yet, to what extent the potentially unlimited growth of co-deposited layers and the corresponding tritium in-vessel inventory pose a problem for the operation of ITER with respect to the maximum in-vessel inventory permitted by radiation safety regulations [1]. The tritium co-deposition problem is of particular concern with respect to the possible migration of material into gaps of macro brush structures or gaps between tiles where it can redeposit at side faces hidden from direct access.…”

Section: Introductionmentioning

confidence: 99%

See 2 more Smart Citations

Formation of deuterium–carbon inventories in gaps of plasma facing components

Krieger

Jacob

Rudakov³

et al. 2007

Journal of Nuclear Materials

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Plasma facing components for ITER will be manufactured as macro brush structures or with castellated surfaces. Material samples with gaps of similar geometry as intended for ITER were exposed to different plasma conditions in TEXTOR, DIII-D and ASDEX Upgrade. In all devices a decrease of both carbon and deuterium inventories at the side faces from the gap entrance into the gap with scale-lengths in the mm range is found. The fraction of D retained at the gap surfaces is in the range of 0.4-4% of the incident flux. Main parameters determining the retained D-fraction are the temperature of the respective surfaces and the carbon fraction in the incident flux. Extrapolation of tritium inventory growth rates to ITER dimensions assuming the measured retention fractions at T>200ºC and using a D/T-flux distribution with a carbon fraction of from B2/EIRENE simulations of an ITER H-mode discharge yields 1% ≈ 1 a contribution to the increase of the total in-vessel tritium inventory in the range of 0.5-5g T/discharge.

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Contribution of the Different Erosion Processes to Material Release from the Vessel Walls of Fusion Devices during Plasma Operation

Behrisch

2002

Contrib. Plasma Phys.

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In high temperature plasma experiments several processes contribute to erosion and loss of material from the vessel walls. This material may enter the plasma edge and the central plasma where it acts as impurities. It will finally be re‐deposited at other wall areas. These erosion processes are: • evaporation due to heating of wall areas. At very high power deposition evaporation may become very large, which has been named “blooming”. Large evaporation and melting at some areas of the vessel wall surface may occur during heat pulses, as observed in plasma devices during plasma disruptions. • At tips on the vessel walls and/or hot spots on the plasma exposed solid surfaces electrical arcs between the plasma and the vessel wall may ignite. They cause the release of ions, atoms and small metal droplets, or of carbon dust particles. • Finally, atoms from the vessel walls are removed by physical and chemical sputtering caused by the bombardment of the vessel walls with ions as well as energetic neutral hydrogen atoms from the boundary plasma. All these processes have been, and are, observed in today's plasma experiments. Evaporation can in principle be controlled by very effective cooling of the wall tiles, arcing is reduced by very stable plasma operation, and sputtering by ions can be reduced by operating with a cold plasma in front of the vessel walls. However, sputtering by energetic neutrals, which impinge on all areas of the vessel walls, is likely to be the most critical process because ions lost from the plasma recycle as neutrals or have to be refuelled by neutrals leading to the charge exchange processes in the plasma. In order to quantify the wall erosion, “Materials Factors” (MF) have been introduced in the following for the different erosion processes.

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Assessment of erosion and tritium codeposition in ITER-FEAT

Cited by 131 publications

References 19 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Formation of deuterium–carbon inventories in gaps of plasma facing components

Contribution of the Different Erosion Processes to Material Release from the Vessel Walls of Fusion Devices during Plasma Operation

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