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.
Consultancy Meeting on Steel Surfaces: Wednesday 20 August 2014 at IAEA Headquarters in Vienna. Meeting report and presentations.
First Research Coordination Meeting: Wednesday-Friday 09-11 December 2015 at IAEA Headquarters. Meeting report and Agenda and presentations.
Second Research Coordination Meeting: Monday-Wednesday 16-18 October 2017 at IAEA Headquarters. Meeting report and Agenda and presentations.
Third Research Coordination Meeting: Expected in Q4 2018 or Q1 2019.
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:
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.
Peng Wang and Qiao Li, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences (with Guang-Nan Luo, CAS IPP and Guang-Hong Lu, Beihang University): Interactions between RAFM steel and boundary plasma.
Wolfgang Jacob, Thomas Schwarz-Selinger et al., Max Planck Institute for Plasma Physics (IPP), Garching, Germany: Laboratory measurements of RAFM steel erosion and H retention in RAFM steel.
Yoshihiko Hirooka and Naoko Ashikawa, National Institute for Fusion Science (NIFS), Toki, Japan: Bi-directional hydrogen isotopes permeation through the first wall of a magnetic fusion power reactor.
Anna Golubeva and Alexander Spitsyn, Fusion Reactor Department, Division of Tokamak Physics, NRC Kurchatov Institute: Retention and permeation of hydrogen isotopes through RAFMs.
Russel Doener and Daisuke Nishijima, University of California at San Diego: Study of PMI with Steel Surfaces in PISCES.
Vadim Maklai, Igor Garkusha et al., IPP Kharkov: Modification of steel surfaces exposed by hydrogen/helium plasmas streams simulating fusion reactor conditions.
Dmitry Terentyev and Lorenzo Malerba, SCK-CEN, Mol, Belgium: Interaction of high flux plasma with Reduced Activation ferritic steels: experimental and computational assessement in baseline RAFM Eurofer 97 and its advanced grades improved by thermomechanical treatment.
General information about participation in IAEA CRPs may be found at the Coordinated Research Activities web site.
Journal literature to follow.
Recent conference contributions relevant to the CRP on Steel Surfaces.
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.