CRP(2015-2020): Plasma-wall Interaction with Reduced-activation Steel Surfaces in Fusion Devices
Various kinds of reduced-activation steel are being considered as wall material for a fusion reactor, but not enough is known about plasma interaction, erosion and tritium retention in such steels. Erosion brings impurities into the plasma and limits the lifetime of the wall. Hydrogen penetration and retention in the surface removes tritium from the plasma, making it unavailable for fusion. Erosion and hydrogen retention properties are affected by damage in the wall material due to neutrons and energetic plasma particles. The CRP on steel surfaces will enhance the knowledge base and develop new databases on interaction of fusion plasma with reduced-activation steel alloys that are considered for fusion. The CRP will seek to quantify the erosion due to exposure to plasma and to quantify the retention and transport properties of tritium in the surface.
Plasma-wall interaction (PMI) issues of erosion and tritium retention are serious problems for fusion energy production. The wall material must be able to withstand a high heat load and particle load and it must have a low affinity for hydrogen (tritium). The present thinking about a fusion demonstration experiment, DEMO, or a reactor is that tungsten will be used in the regions of highest heat load, but some kind of reduced-activation steel could be preferred for the main wall. (The main concerns with tungsten are radiation embrittlement, increased affinity for hydrogen after radiation damage and very high power loss due to tungsten impurities in the main plasma.) However, the PMI properties of steel after irradiation, the resistance to sputtering and ablation, and the behaviour of trapped tritium in steel surfaces in a fusion neutron environment are not well enough known at present to make an informed assessment about the possible role of steel as a plasma facing material.
The CRP on "Plasma-wall interaction with reduced-activation steel surfaces in fusion devices" is intended to enhance the knowledge base on erosion, tritium deposition and tritium migration processes involving fusion relevant (reduced activation) steel surfaces. The plasma-wall interaction processes include sputtering by H and He and plasma impurities, trapping of hydrogen (H, D, T) in surfaces exposed to plasma, transport of hydrogen in the steel and means to extract trapped tritium. The CRP will bring together experimentalists from fusion research institutes and from laboratory plasma-material interaction experiments and theorists involved in applied studies of plasma and neutron interaction with steels.
The CRP on Steel Surfaces is meant to support planning and design efforts in Member States towards DEMO and a Fusion Power Plant through the enhancement of the knowledge base on properties of reduced-activation steels as a plasma-facing material in a fusion nuclear environment. Important properties include the relation between steel microstructure and erosion, hydrogen retention and hydrogen migration. More specific objectives are:
- To perform investigations and assemble information about the characterization of microstructure of steel surfaces exposed to fusion neutrons and energetic plasma particles.
- To perform investigations and assemble information about the relation between steel microstructure after irradiation and plasma-material interaction properties for erosion, tritium retention and tritium migration.
- To perform investigations and assemble information about ways to mitigate tritium penetration and tritium retention in steel surfaces and to extract trapped tritium.
- To synthesize new information and provide best expert estimates and uncertainties for plasma-material interaction properties (especially tritium retention and tritium transport) for steel surfaces in a fusion reactor environment.
Finally the CRP will lead to increased confidence in assessments of the role of steel as a plasma-facing material in DEMO or in a Fusion Power Plant.
Journal literature to follow.
CRP F44003 on "Primary Radiation Damage Cross Sections" (2013-2017) engages participants from the nuclear data and materials research communities to determine ways to characterize in a concise quantitative way radiation damage in materials by neutrons and charged particles. In the CRP on steel surfaces the radiation damage needs to be correlated with tritium retention and tritium transport properties.
CRP F43020 on "Data for erosion and tritium retention in beryllium plasma-facing materials" (2012-2017) is concerned with erosion due to physical and chemical sputtering of beryllium, trapping and reflection of hydrogen (H, D, T) on beryllium surfaces, the transport of hydrogen in beryllium and means to extract trapped tritium. The CRP on steel surfaces is concerned with the same class of processes, but in the context of new experiments in which low-activation steel takes the place of beryllium as wall material.
CRP F43021 on "Plasma-wall interaction for irradiated tungsten and tungsten alloys in fusion devices" (2013-2018) is intended to enhance the knowledge base on the effects of fusion relevant irradiation on plasma-wall interaction and hydrogen retention in tungsten-based materials. Tungsten remains the favoured material for wall regions of highest heat load; therefore the CRP on irradiated tungsten and the one on steel surfaces together span the presently favoured wall materials for a fusion nuclear device.