International Doctoral College in Fusion Science and Engineering
 Thesis catalogue
Study of retention and release of Helium and their mechanisms in Beryllium and Tungsten Plasma Facing Components, application to ITER
PhD Code: 2016-DC-16:
  • Host institute 1: AM07-Aix- Marseille Université (Home University) - FP8-Institut de Recherche sur la Fusion par confinement magnétique, Saint-Paul-lez-Durance, France (Home Institution)
  • Host institute 2: AM26- Technische Universität München (Host University) - FP9-Max-Planck-Institut für Plasmaphysik Garching und Greifswald (Host Institution)
Research fields:
  • F4. Plasma-wall interaction and material research
  • Prof. Jean-Marc Layet (promotor) - Dr. David Douai (mentor)
  • Prof. Rudolf Neu (co-promotor)
Contact Person and email: Douai David -

Subject description
Background: ITER, currently in construction in Cadarache, France, will be the largest tokamak aiming at demonstrating the feasibility of thermonuclear fusion by magnetic confinement ( ITER will have a Beryllium (Be) first wall and a Tungsten (W) divertor. In the nuclear phase of ITER, these Plasma facing Components (PFCs), in particular the divertor, which will have to handle the highest heat and particle loads, will be exposed to high fluences of Deuterium, Tritium or Helium ashes coming from the fusion reaction. Helium operation is also foreseen in the non-nuclear phase of ITER. Bombardment by He plasma particles will result in different consequences regarding the lifetime of PFCs, the operation or the performance of fusion plasmas in ITER. Erosion of Be surfaces at the first wall, followed by fuel implantation in Be and W and co-deposition with Be in the divertor region, is known to be the main cause of fuel retention in ITER. But, while retention and release of hydrogenic fuel has been extensively studied by experiments, in laboratory and in tokamaks, in particular JET with the ITER-Like Wall (Be/W), as well as by modelling, the database on the interaction of Helium with Be and W is much less detailed. Interaction between Helium and metallic PFCs, even at modest fluences, is also known to induce specific structural changes, like the formation of W nanostructures, or of He nanobubbles, observed in both Be and W, impacting fuel recycling properties and degrading thermal properties of PFCs.

Expected outcomes

Objective: The aim of this thesis is to study the interaction between He plasmas and Be and W PFCs in order to document and predict possible consequences for the operation of ITER. The work will consist in participating in dedicated He or He/D plasma experiments (tokamak plasma and conditioning discharges) in current devices with metallic PFCs (Tore Supra-WEST in Cadarache, ASDEX-Upgrade in Germany), and analysing data in order to assess Helium and hydrogenic fuel and retention and release. In parallel, the student will take part in the definition and the analysis of He (He/D) plasma surface (Be/W) interaction experiments in laboratory, under well-defined conditions, representative of those encountered in tokamak or wall conditioning plasmas in terms of energy, flux and fluence of impinging ions, surface state (e.g. with implanted or co-deposited hydrogenic fuel with Be, as observed in JET-ILW) or surface temperature. These laboratory experiments will take place in IPP for Tungsten related work, while for the Beryllium related work, a collaboration with the PISCES facility (US) is foreseen. In order to understand the interaction mechanisms, a diffusion-trapping model for He in Be/W surfaces should be developed on the basis of existing models, and the results compared with experiments. Finally, this work, combining experiments in fusion devices and laboratory, and modeling, should aim at consolidating understanding on retention and release in presence of Helium, to identify the main plasma wall interaction mechanisms at play, and to predict consequences for the operation of ITER.


Time line and mobility scheme (research need to be performed for at least six month in two different countries): 1st year (October 2016-September 2017): M1-6: IRFM Understanding of relevant PWI issues in tokamaks, in particular in ITER, bibliographic study. M7-9: definition of laboratory experiments. M10-12: participation in first dedicated He experiments on WEST 2nd year: (October 2017-September 2018): M1-3 : participation in dedicated laboratory experiments at IPP M4-6 : participation in dedicated He experiments on AUG M7-9 : participation in dedicated He experiments on WEST M10-12 : data analysis & modeling 3rd year (October 2018-September 2019): M1-3 : data analysis & modeling, participation in complementary laboratory experiments at IPP M4-6 : data analysis & modeling, M7-12 : IRFM Write up of thesis and submission



Orignial document: 2016-DC-16

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