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Plasma-Wall Interaction (PWI)

The field of plasma-wall interaction (PWI) comprises all processes involved in the exchange of mass and energy between plasma and the surrounding wall materials.

This broad interdisciplinary field of is essential for interrelated aspects of fusion reactor operation: economy and safety. To be tackled are the following essential issues:

  • Lifetime of plasma-facing materials (PFM) and components (PFC)
  • Accumulation of hydrogen isotopes in PFC, i.e. tritium inventory
  • Carbon and metal (Be, W) dust formation

PWI is one of the primary areas where integration of the physics and technology programmes is being achieved. Our work in fusion-related material physics has been fully integrated with:

  1.  EU Fusion Programme
  2.  International Tokamak Fusion Activity (ITPA) under auspices of ITER
  3. Implementing Agreements of International Energy Agency (IEA).

It is demonstrated by active participation (proposing and initiating tasks) and responsibilities for research projects and coordination in:

  • EUROfusion (formerly known as EFDA) Work Programme on ITER Physics and Power Plant Physics and Technology
  • EUROfusion-JET Work Programme: Experimental campaigns and coordination of studies of plasma-facing components from JET
  • ITPA on the (a) Divertor and Scrape-off Layer Physics and (b) Diagnostics
  • IEA activities
  • International Union for Vacuum Science, Technique and application (IUVSTA)
  • Organisation of conferences and edition of proceedings in international journals.

Goal

The strategic goal is to carry out a strong experimentally-based program focused on ITER-relevant topics and exploiting existing devices in the European program:

  Specific aim
a.  The assessment of material lifetime and processes of dust formation
b.  The determination of material performance in operation with metal wall (Be and W) 
c.  The determination of long-term fuel inventory in PFC
d.  Contribution to development and testing of fuel removal methods
e.  Development of in-vessel diagnostics for material migration studies
f. Development and application of surface analysis methods for studies of plasma-facing components

Highlights of recent activities and achievements

  Highlight
a.  Detailed characterisation of erosion/deposition scenario in JET with all consecutive divertors bot in JET with carbon wall (JET-C) and JET-ILW with a metal wall: material distribution quantified and mapped; the results for JET-C have been modelled.
b. Tracer technique (based on injection of marker gases [13-C labelled methane (13CH4), Nitrogen-15 and Oxygen-18]) has been further developed and used at TEXTOR, ASDEX and JET to determine material migration and preferential flows in the scrape-off layer (SOL).  The results have been modelled.
c. Beryllium transport in JET has been determined. 
d. A set of erosion-deposition diagnostics for JET-ILW has been developed, tested and installed in order to study material migration.
e. Very detailed survey of dust in JET-ILW was accomplished. 
f. For the first time long-term fuel inventory and co-deposition in castellated beryllium limiters and divertor structures were quantified. The results clearly show the decisive role of carbon transport on the fuel inventory in castellated metal PFC. (Castellation: structure like a chocolate bar with narrow gaps between segments. All PFC in ITER will be castellated.) 
g. Beryllium marker tiles and coatings for application at ILW-JET were developed and installed in JET-ILW. Studies of markers retrieved from JET-ILW are under way.
h. Comprehensive characterisation of different PFC from Wendelstein 7-X and WEST was performed.
i. Efficacy of fuel removal and surface state after cleaning were determined for four techniques being developed for detritiation: photonic cleaning by flash lamp, laser-induced detritiation, oxygen- and nitrogen-assisted discharges. Recently efforts have been concentrated on Ion Cyclotron and Electron Cyclotron Wall Conditioning (ICWC and ECWC, respectively) for fuel control in wall components.
j. Plasma impact on High-Z metals has been studied. No fuel accumulation in tungsten was detected. Experimental work has been integrated with modelling. 
k. The First Mirror Test for ITER was completed in JET-C and has been continued in JET-ILW. Until now it has been the most comprehensive mirror test performed in a tokamak. 
l. Nuclear microprobe has been developed and broadly used for studies of wall materials from tokamaks.

The results obtained have had significant implications for the progress in fusion science and technology, in particular:

  • Planning of a large-scale fusion experiment: ITER-Like Wall Project at JET
  • Selection and shaping of PFM in future devices
  • Identification of conditions governing transport and fuel retention in remote areas of the divertor and in castellated PFC

Publications: over 300 papers in international journals

Cooperation

Collaborator (proven by joint publication, conference contributions)
Joint European Torus (JET) at the Culham Science Centre, United Kingdom
Ångström Laboratory, Uppsala University: ion beam analysis of PFC
Forschungszentrum Jülich (FZJ), Germany: TOMAS plasma facility for the development of fuel removal methods
Max-Planck-Institute of Plasma Physics, Garching, Germany  
The Royal Military Academy, Brussels, Belgium
International Fusion Energy Research Centre (IFERC), Rokkasho, Japan
Institute of Radiation, Laser and Atomic Physics, Bucharest-Magurele, Romania
WEST tokamak at CEA, Cadarache, France
Plasma physics and materials science laboratories in Finland, Poland, Croatia, Greece, Portugal
Institute of Plasma Physics, Czech Academy of Sciences: COMPASS tokamak

Members of the research group

Innehållsansvarig:Björn Zaar
Tillhör: Fusionsplasmafysik
Senast ändrad: 2021-08-31