Tritium depth profiles in 2D and 4D CFC tiles from JET and TFTR

Bekris, N.; Skinner, C.H.; Berndt, U.; Gentile, C.; Glugla, M.; Schweigel, B.

doi:10.1016/s0022-3115(02)01576-3

Cited by 29 publications

(56 citation statements)

References 14 publications

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“…However, it is also possible that deuterium/tritium could be trapped inside the CFC itself. This hypothesis seems to be corroborated by results from some laboratories [13] but also by Bekris [22] who has observed deep diffusion of tritium in JET CFC tiles. Tritium was detected at a substantial level even at the centre of a tile with a thickness of some centimetres.…”

Section: Consequences Of Bulk Materials Tritium Trappingsupporting

confidence: 61%

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Treatment of ITER plasma facing components: Current status and remaining open issues before ITER implementation

Grisolia¹,

Counsell²,

Dinescu³

et al. 2007

Fusion Engineering and Design

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The in-vessel tritium inventory control is one of the most ITER challenging issues which has to be resolved to fulfil safety requirements. This is due mainly to the presence of carbon as a constituent of plasma facing components (PFCs) which leads to a high fuel permanent retention. For several years now, physics studies and technological developments have been undertaken worldwide in order to develop reliable techniques which could be used in ITER severe environment (magnetic field, vacuum, high temperature) for in situ tritium recovery. The scope of this contribution is to review the present status of these achievements and define the remaining work to be done in order to propose a dedicated work program.Different treatment techniques (chemical treatments, photonic cleaning) will be reviewed. In the frame of ITER, they will be compared in terms of fuel removal rate as well as surface accessibility, type of production (gas or particulates), ability to clean mixed material.And lastly, consequences of bulk trapping observed in tokamak on the techniques currently under development will be addressed.Published with Fusion Engineering and Design: Fus. Eng. Des. 82 (2007Des. 82 ( ) 2390Des. 82 ( -2398

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Section: Consequences Of Bulk Materials Tritium Trappingsupporting

confidence: 61%

“…Detritiated zones will be characterised using ion beam analysis [21], calorimetry [22] and combustion of cored samples [23].…”

Section: A Laboratory and Behf Assessmentsmentioning

confidence: 99%

Treatment of ITER plasma facing components: Current status and remaining open issues before ITER implementation

Grisolia¹,

Counsell²,

Dinescu³

et al. 2007

Fusion Engineering and Design

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“…A minor proportion of tritium was found on plasma facing surfaces and most of it was retained in co-deposited C-layers at the inner divertor. Only a small fraction of the tritium diffused deeper inside the material along the open porous structure of the CFC material [421].…”

Section: Database On Fuel Retention In Present Fusion Devicesmentioning

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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“…Because of its high thermal shock resistivity and high thermal conductivity, carbon material is a candidate as a divertor target in ITER. Carbon materials are easily eroded by incident hydrogen bombardment and the sputtered carbon atoms are co-deposited with hydrogen isotopes on the walls nearby the divertor target [2][3][4][5][6]. Hydrogen isotopes retention in carbon films retained by co-deposition and implantation processes have been well investigated so far [7,8].…”

Section: Introductionmentioning

confidence: 99%