Material erosion and migration in tokamaks

The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the susbsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasmafacing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined.

Chapter 4: Power and particle control

Richard Pitts

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.

2007

Overview of results and possibilities for fast particle research on JET

Olivier Sauter, Duccio Testa

The large physical size of the JET tokamak, its heating systems and diagnostics, and its capability to operate with full deuterium-tritium (D-T) plasmas, including high-power tritium neutral beam injection (NBI), give it unique possibilities in fast particle research in fusion plasmas. These have already been used to generate significant (2-3 MW level) power in fusion a-particles in the 1997 D-T campaign. Recent JET experiments have concentrated on two important scenarios of relevance to next-step tokamak devices: the ELMy H-mode plasmas and plasmas with strong internal transport barriers (ITBs). The achieved progress will help in preparation for a possible second D-T experiment on JET. Fast particle studies have also been carried out recently using ion cyclotron resonance heating (ICRH)-accelerated particles and external-excitation methods to study Alfven eigenmodes (AEs). Looking towards the future, the capability of JET will be enhanced by upgrades to the NBI system, ICRH system and various diagnostics. Results of the first JET D-T experiment (DTE1) form a basis on which to elaborate a second D-T experiment (DTE2) which could be proposed after these enhancements. The alpha-physics part of this programme would be divided between the investigation of alpha-particle confinement, heating and loss processes in the 'integrated scenarios' (where the discharge is as close as possible to an ITER-relevant scenario), and dedicated 'alpha-physics' experiments, with specially prepared plasmas. In ELMy H-mode plasmas the fusion performance could roach Q(=P-fusion/P-input) of similar to0.33 at the highest combined heating powers, corresponding to similar to 6x10(-4), allowing a test of the margins of TAE stability in quasi-steady-state conditions. The integrated-scenario fast particle programme could concentrate on the instabilities and heating in plasma regimes with strong steady-state ITBs, with expected Q values similar to0.58 and similar to2x10(-3), demonstrating the compatibility of these operating scenarios with alpha-effects. Excitation of TAEs by alpha-particles in the plasma core could also be studied in such integrated scenarios. An issue which will receive attention is the confinement of MeV energy ions in the centre of ITB plasmas with strongly reversed shear, where the low current density in the centre may lead to the alpha-particles entering loss orbits. In preparation for a D-T campaign, studies of triton burn-up in deuterium ITB plasmas will begin in the 2002 experimental campaigns. Special 'afterglow' experiments to measure TAEs after the termination of the (stabilizing) NBI have already been explored in JET deuterium ITB scenarios and would be planned for DTE2. It is intended to develop special versions of ITB plasmas with dominant ion heating which would maximize the sensitivity to degradation of alpha-heating effects.

2002

A fast-moving Langmuir probe array for the divertor of the Tokamak à Configuration Variable

Hugo De Oliveira

In next step tokamaks such as ITER and future fusion power plants, wall materials will be subjected to considerably higher heat and particle fluxes than in present-day machines. Maintaining plasma facing components in good condition and simultaneously assuring acceptable core performance will be an outstanding challenge. This situation is further complicated by the fact that extrapolations towards future tokamaks are difficult and the mechanisms governing transport of energy and particles are still being investigated. Proper diagnostic coverage of the boundary plasma of currently operational machines is therefore crucial. In the Tokamak à Configuration Variable (TCV), wall-embedded Langmuir probes (LPs) are one of the most powerful diagnostics to describe local properties in the boundary region of the plasma. The present thesis work includes operation, maintenance and upgrades of this LP system. In particular, the TCV wall probe coverage has been extended from 114 to 194 probes. Furthermore, a new set of amplifiers has been installed, featuring a number of improvements. In parallel, a novel fast probe array, the Reciprocating Divertor Probe Array (RDPA), has been designed, built and commissioned at TCV, providing unprecedented Langmuir probe measurements across the divertor volume. In contrast to reciprocating probes accessing the divertor region of other major tokamaks, the RDPA provides 2D measurements in a single reciprocation, enabled by a radial array of Mach probe tips moved vertically across an extended region of the divertor volume. The RDPA capabilities have allowed for a detailed particle balance study in the outer TCV divertor in attached and detached conditions and both with and without the recently installed divertor baffles. It has been observed that the contribution of the ExB drift to the poloidal ion flux is comparable and sometimes larger than that due to the ion flux parallel to the magnetic field. Except for strongly detached conditions, most of the particle flux arriving at the outer target is shown to be due to ionization along the outer divertor leg, justifying the 'closed-box approximation' frequently used in detachment models and previously inferred on TCV from spectroscopic measurements. ExB drifts also play an important role in terms of heat transport, as found from a divertor heat flux balance study in attached conditions. Together with parallel heat conduction, the ExB-convected heat flux dominates the poloidal heat flux, while parallel heat convection plays a lesser role in the example studied. A new feature of RDPA is also to provide detailed, local measurements of fluctuation characteristics across the divertor volume of TCV, thanks to acquisition frequencies of up to 2MHz. The first findings include the development of a strong poloidal gradient of the density fluctuation level, observed towards the onset of detachment, and the surprising presence of high fluctuation levels in the narrow region of downward ExB (on the high-field side of the peak electron temperature) for well attached discharges. Ongoing efforts focus on using RDPA results to validate the global 3D two-fluid code GBS in diverted TCV simulations and thus helping to improve its predictive capabilities.

EPFL2021