The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), 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 scrape-off layer (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, edge localized mode (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 in 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 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.

/pdf/chapter-4-power-and-particle-control-3l68lj0fxw.pdf

Chapter 4: Power and particle control

Structural materials for fission & fusion energy

A full tungsten divertor for ITER: Physics issues and design status

Tungsten as material for plasma-facing components in fusion devices

Plasma facing components in future thermonuclear fusion devices will be subjected to intense transient thermal loads due to type I edge localized modes (ELMs), plasma disruptions, etc. To exclude irreversible damage to the divertor targets, local energy deposition must remain below the damage threshold for the selected wall materials. For monolithic tungsten (pure tungsten and tungsten alloys) power densities above ?0.3?GW?m?2 with 1?ms duration result in the formation of a dense crack network. Thin tungsten coatings for the so-called ITER-like wall in JET, which have been deposited on a two-directional carbon?fibre composite (CFC) material, are even less resistant to thermal shock damage; here the threshold values are by a factor of 2 lower. First ELM-simulation experiments with high cycle numbers up to 104 cycles on actively cooled bulk tungsten targets do not reveal any cracks for absorbed power densities up to 0.2?GW?m?2 and ELM-durations in the sub-millisecond range (0.8?ms); at somewhat higher power densities (0.27?GW?m?2, ?t = 0.5?ms) cracks have been detected for 106 cycles.

https://iopscience.iop.org/article/10.1088/0029-5515/51/7/073017/pdf

Performance of different tungsten grades under transient thermal loads

Neutron irradiation effects on plasma facing materials

Qualification and Post-Mortem Characterization of Tungsten Mock-ups Exposed to Cyclic High Heat Flux Loading

High heat flux testing of plasma facing materials and components – Status and perspectives for ITER related activities

Beside quasi-stationary plasma operation, short transient thermal pulses with deposited energy densities on the order of several tens of MJ/m2 are a serious concern for next-step devices, in partic...

M. Rödig

Papers

Performance of different tungsten grades under transient thermal loads

Neutron irradiation effects on plasma facing materials

Qualification and Post-Mortem Characterization of Tungsten Mock-ups Exposed to Cyclic High Heat Flux Loading

High heat flux testing of plasma facing materials and components – Status and perspectives for ITER related activities

Performance of Plasma-Facing Materials Under Intense Thermal Loads in Tokamaks and Stellarators