Material/plasma surface interaction issues following neutron damage

Barabash, V.; Federici, G.; Linke, J.; Wu, C.H.

doi:10.1016/s0022-3115(02)01330-2

Cited by 79 publications

(48 citation statements)

References 40 publications

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“…[334] or the full erosion dependence on energy, surface temperature and particle flux [276]. Calculations are done for a high and a low surface temperature case [412], (after [333]). …”

Section: Effects On Tritium Removal Techniques (Oxidation or Heating)mentioning

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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“…[334] or the full erosion dependence on energy, surface temperature and particle flux [276]. Calculations are done for a high and a low surface temperature case [412], (after [333]). …”

Section: Effects On Tritium Removal Techniques (Oxidation or Heating)mentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…The base temperature was taken from Fig. 5 in [22]. The maximum temperature was limited to 1500K since above this temperature Be and some of the Be/W alloys become liquid [23].…”

Section: Influence Of Temperature Excursionsmentioning

confidence: 99%

Beryllium flux distribution and layer deposition in the ITER divertor

Schmid

2008

Nucl. Fusion

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Abstract.The deposition of Be eroded from the main chamber wall on the W surfaces in the ITER divertor could result in the formation of Be rich Be/W mixed layers with a low melting temperature compared to pure W. To predict whether or not these layers form the Be flux distribution in the ITER divertor is required. This paper presents the results of a combination of plasma transport-with erosion/deposition simulations that allow to calculate both the Be flux distribution and the Be layer deposition in the ITER divertor. This model includes the Be source due to Be erosion in the main chamber and the deposition, re-erosion and re-deposition of Be in the ITER divertor. The calculations show that the fraction of Be in the incident particle flux in the divertor range from ≈ 10 −3 to ≈ 5% with a pronounced inner outer divertor asymmetry. The flux fractions in the inner divertor are on average ten times higher than in the outer divertor. Thick Be layers only from at the inner strike point and the dome baffles. The highest Be layer growth rate is found to be 1.0 nm/s. Despite the Be deposition the formation of Be rich Be/W mixed layers is not to be expected in ITER. The expected surface temperature at these locations during steady state operation is too low as to result in Be diffusion into W and thus Be/W mixed layers can not form.

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“…In a further step, selected material samples and small-scale modules were irradiated in material test reactors to ITER specific fluences [4]. In a final step, medium and full-scale components have been manufactured by the European industry.…”

Section: Plasma Facing Component Developmentmentioning

confidence: 99%

“…Carbon based materials show a rather strong decrease in thermal conductivity even after relatively low neutron fluences [4]. The ITER candidate CFC armour material NB31 for example exhibits excellent thermal conductivities before neutron irradiation.…”

Section: Materials Performance After Neutron Irradiationmentioning

confidence: 99%