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Based on the article by Wilson et al. in APID vol. 1 (1991)[1] and in Data Compendium for Plasma-Surface interactions (1984) [2]

As energetic hydrogen atoms and ions impinge on a solid surface, some of the incident flux will be directly reflected, the remaining (penetrating) flux will thermalise inside the solid and will either become trapped or begin to diffuse in the lattice. Hydrogen that reaches the surface can be released as atoms by photon, electron, or ion impact desorption or as molecules by thermally activated molecular recombination. Thermal release of hydrogen atoms from a solid is not thermodynamically favored because of the high dissociation energy (4.5 eV) of the hydrogen molecule; thus, thermal release does not become significant until the temperatures reach ~ 2000 K.

Hydrogen isotope trapping and release by plasma facing materials control predominantly fuel retention and recycling in magnetic confinement fusion devices. Plasma-wall interactions involving hydrogenic particles have a major influence not only on the lifetime of the first wall and other plasma facing components but also on the fundamental properties of heated plasmas. In future D-T reactors, the hydrogen transport and desorption behaviour will control the tritium inventory in first walls and limiter/divertors and the plasma-driven tritium permeation into the coolant.

Hydrogen trapping and release in fusion materials have been studied extensively in the laboratory and in large scale fusion experiments. In parallel, a phenomenological theory based on simple diffusion in the presence of bulk and surface defects has been developed to account for the transport of hydrogen isotopes in and out of materials during and after exposure to hydrogenic plasmas.

In recent years, it has been endeavoured to reduce Zeff in fusion experiments and, consequently, there has been increased use of low-Z refractory plasma facing materials such as carbon and beryllium. However, in the course of the ITER design process the use of high-Z materials such as molybdenum or tungsten is considered in cases where a low plasma edge electron temperature can be maintained in a high recycle divertor.

Trapping and release of hydrogen in metals

Hydrogen retention and release characteristics for Beryllium

Hydrogen retention and release characteristics for Graphite

Hydrogen Behaviour in Molybdenum and Tungsten

Bulk diffusion, solubility and trapping of hydrogen and graphite at elevated temperatures

Detritiation techniques

Research Activities

IAEA conducted a cooredinated research project on Tritium Inventory in Fusion Reactors in the period of 2002 to 2006. The summary of the CRP research is published in a joint paper[3] in Fusion Science and Technology and a APID volume. Presentations on the following topics can be found in the IAEA A+M Homepage[4].

More information on the related research acitivities by institutes worldwide can be found:


  1. R. K. Janev and A. Miyahara , Plasma-material interaction issues in fusion reactor design and status of the database, Atomic and Plasma-Material Interaction Data for Fusion, v.1 p.123 (1991)
  2. R. A. Langley, J. Bohdansky, W. Eckstein, P. M ioduszewski, J. Roth, E. Taglauer, E. W. Thomas, H. Verbeek, K. L. Wilson, Data Compendium for Plasma-Surface interactions, Nucl. Fusion, Special Issue 1984, IAEA, Vienna
  3. C.H. Skinner, A.A. Haasz, V.Kh. Alimov, N. Bekris, R.A. Causey, R.E.H. Clark, J.P. Coad, J.W. Davis, R.P. Doerner, M. Mayer, A. Pisarev, J. Roth, T. Tanabe, Recent Advances on Hydrogen Retention in ITER's Plasma-Facing Materials: Beryllium, Carbon, and Tungsten, Fusion Science and Technology 54 (2008) 891-945
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