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.

/pdf/plasma-material-interactions-in-current-tokamaks-and-their-1a6iwv89x6.pdf

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

http://ieeexplore.ieee.org/iel5/4639142/31561/01472527.pdf

Controlled fusion

Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which made possible the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in studying the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation that both have good confinement and are MHD stable has made possible a broad study of problems in burning-plasma physics. The technical and scientific results from deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) are reviewed, with particular emphasis on alpha-particle physics issues.

/pdf/results-from-deuterium-tritium-tokamak-confinement-161dumql17.pdf

Results from deuterium-tritium tokamak confinement experiments

Tritium retention of plasma facing components in tokamaks

Monte Carlo neutral transport simulations of hydrogen velocities in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] are compared with experiment using the Doppler‐broadened Balmer‐α spectral line profile. Good agreement is obtained under a range of conditions, validating the treatment of charge exchange, molecular dissociation, surface reflection, and sputtering in the neutral gas code DEGAS [D. Heifetz et al., J. Comput. Phys. 46, 309 (1982)]. A residual deficiency of 10–100 eV neutrals in most of the simulations indicates that further study of the energetics of H+2 dissociation for electron energies in excess of 100 eV is needed.

https://publish.illinois.edu/plasma-material-interactions/files/2015/04/1996StotlerSkinnerBudnyRamsey-RuzicTurkot.pdf

Modeling of neutral hydrogen velocities in the Tokamak Fusion Test Reactor

The particle throughput of TPX must be greater than 85 Torr-l/s to ensure the capability of steady-state operation. This throughput must be maintained by active pumping and clever design of the divertor components. In this paper seven geometrical variations of the divertor structures are modeled using the B2 edge-plasma fluid code and DEGAS Monte-Carlo neutral transport code. Results show that eliminating the gap between the inner divertor plate and the baffle, and reducing the gap between the outer divertor plate and the baffle by two-thirds will improve the particle throughput over the reference design by a factor of four. The resulting throughput from this new geometry is 91.8/spl plusmn/10.9 Torr-l/s.

R. B. Turkot

Papers

Modeling of neutral hydrogen velocities in the Tokamak Fusion Test Reactor

TPX divertor design: neutral atom behavior and optimization for pumping